INTEGRATED PLASMA CONTROL FOR ADVANCED TOKAMAKS
HUMPHREYS,D.A; FERRON,J.R; JOHNSON,R.D; LEUER,J.A; PENAFLOR,B.G; WALKER,M.L; WELANDER,A.S; KHAYRUTDINOV,R.R; DOKOUKA,V; EDGELL,D.H; FRANSSON,C.M
2003-10-01
OAK-B135 Advanced tokamaks (AT) are distinguished from conventional tokamaks by their high degree of shaping, achievement of profiles optimized for high confinement and stability characteristics, and active stabilization of MHD instabilities to attain high values of normalized beta and confinement. These high performance fusion devices thus require accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating, as well as simultaneous and well-coordinated MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Satisfying the simultaneous demands on control accuracy, reliability, and performance for all of these subsystems requires a high degree of integration in both design and operation of the plasma control system in an advanced tokamak. The present work describes the approach, benefits, and progress made in integrated plasma control with application examples drawn from the DIII-D tokamak. The approach includes construction of plasma and system response models, validation of models against operating experiments, design of integrated controllers which operate in concert with one another as well as with supervisory modules, simulation of control action against off-line and actual machine control platforms, and iteration of the design-test loop to optimize performance.
Advanced Tokamak Plasmas in the Fusion Ignition Research Experiment
C.E. Kessel; D. Meade; D.W. Swain; P. Titus; M.A. Ulrickson
2003-10-13
The Advanced Tokamak (AT) capability of the Fusion Ignition Research Experiment (FIRE) burning plasma experiment is examined with 0-D systems analysis, equilibrium and ideal-MHD stability, radio-frequency current-drive analysis, and full discharge dynamic simulations. These analyses have identified the required parameters for attractive burning AT plasmas, and indicate that these are feasible within the engineering constraints of the device.
ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM
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
2002-10-01
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.
Guo, H. Y.; Li, J.; Wan, B. N. Gong, X. Z.; Xu, G. S.; Zhang, X. D.; Ding, S. Y.; Gan, K. F.; Hu, J. S.; Hu, L. Q.; Liu, S. C.; Qian, J. P.; Sun, Y. W.; Wang, H. Q.; Wang, L.; Xia, T. Y.; Xiao, B. J.; Zeng, L.; Zhao, Y. P.; and others
2014-05-15
A long-pulse high confinement plasma regime known as H-mode is achieved in the Experimental Advanced Superconducting Tokamak (EAST) with a record duration over 30 s, sustained by Lower Hybrid wave Current Drive (LHCD) with advanced lithium wall conditioning and divertor pumping. This long-pulse H-mode plasma regime is characterized by the co-existence of a small Magneto-Hydrodynamic (MHD) instability, i.e., Edge Localized Modes (ELMs) and a continuous quasi-coherent MHD mode at the edge. We find that LHCD provides an intrinsic boundary control for ELMs, leading to a dramatic reduction in the transient power load on the vessel wall, compared to the standard Type I ELMs. LHCD also induces edge plasma ergodization, broadening heat deposition footprints, and the heat transport caused by ergodization can be actively controlled by regulating edge plasma conditions, thus providing a new means for stationary heat flux control. In addition, advanced tokamak scenarios have been newly developed for high-performance long-pulse plasma operations in the next EAST experimental campaign.
NASA Astrophysics Data System (ADS)
Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook
2015-12-01
Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.
Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook
2015-12-01
Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method. PMID:26724028
Kim, Dong-Hwan; Hong, Suk-Ho; Park, Il-Seo; Lee, Hyo-Chang; Kang, Hyun-Ju; Chung, Chin-Wook
2015-12-15
Plasma characteristics in the far scrape-off layer region of tokamak play a crucial role in the stable plasma operation and its sustainability. Due to the huge facility, electrical diagnostic systems to measure plasma properties have extremely long cable length resulting in large stray current. To overcome this problem, a sideband harmonic method was applied to the Korea Superconducting Tokamak Advanced Research tokamak plasma. The sideband method allows the measurement of the electron temperature and the plasma density without the effect of the stray current. The measured plasma densities are compared with those from the interferometer, and the results show reliability of the method.
Profile control of advanced tokamak plasmas in view of continuous operation
NASA Astrophysics Data System (ADS)
Mazon, D.
2015-07-01
The concept of the tokamak is a very good candidate to lead to a fusion reactor. In fact, certain regimes of functioning allow today the tokamaks to attain performances close to those requested by a reactor. Among the various scenarios of functioning nowadays considered for the reactor option, certain named 'advanced scenarios' are characterized by an improvement of the stability and confinement in the plasma core, as well as by a modification of the current profile, notably thank to an auto-generated 'bootstrap' current. The general frame of this paper treats the perspective of a real-time control of advanced regimes. Concrete examples will underline the impact of diagnostics on the identification of plasma models, from which the control algorithms are constructed. Several preliminary attempts will be described.
A long-pulse high-confinement plasma regime in the Experimental Advanced Superconducting Tokamak
NASA Astrophysics Data System (ADS)
Li, J.; Guo, H. Y.; Wan, B. N.; Gong, X. Z.; Liang, Y. F.; Xu, G. S.; Gan, K. F.; Hu, J. S.; Wang, H. Q.; Wang, L.; Zeng, L.; Zhao, Y. P.; Denner, P.; Jackson, G. L.; Loarte, A.; Maingi, R.; Menard, J. E.; Rack, M.; Zou, X. L.
2013-12-01
High-performance and long-pulse operation is a crucial goal of current magnetic fusion research. Here, we demonstrate a high-confinement plasma regime known as an H-mode with a record pulse length of over 30s in the Experimental Advanced Superconducting Tokamak sustained by lower hybrid wave current drive (LHCD) with advanced lithium wall conditioning. We find that LHCD provides a flexible boundary control for a ubiquitous edge instability in H-mode plasmas known as an edge-localized mode, which leads to a marked reduction in the heat load on the vessel wall compared with standard edge-localized modes. LHCD also induces edge plasma ergodization that broadens the heat deposition footprint. The heat transport caused by this ergodization can be actively controlled by regulating the edge plasma conditions. This potentially offers a new means for heat-flux control, which is a key issue for next-step fusion development.
Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma
Xu, Liqing; Zhang, Jizong; Chen, Kaiyun E-mail: lqhu@ipp.cas.cn; Hu, Liqun E-mail: lqhu@ipp.cas.cn; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin; Zhu, Yubao
2015-12-15
Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting mode structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well.
Development of Burning Plasma and Advanced Scenarios in the DIII-D Tokamak
Luce, T C
2004-10-18
Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q {approx} 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque.
Development of Burning Plasma and Advanced Scenarios in the DIII-D Tokamak
Luce, T C
2004-12-01
Significant progress in the development of burning plasma scenarios, steady-state scenarios at high fusion performance, and basic tokamak physics has been made by the DIII-D Team. Discharges similar to the ITER baseline scenario have demonstrated normalized fusion performance nearly 50% higher than required for Q = 10 in ITER, under stationary conditions. Discharges that extrapolate to Q {approx} 10 for longer than one hour in ITER at reduced current have also been demonstrated in DIII-D under stationary conditions. Proof of high fusion performance with full noninductive operation has been obtained. Underlying this work are studies validating approaches to confinement extrapolation, disruption avoidance and mitigation, tritium retention, ELM avoidance, and operation above the no-wall pressure limit. In addition, the unique capabilities of the DIII-D facility have advanced studies of the sawtooth instability with unprecedented time and space resolution, threshold behavior in the electron heat transport, and rotation in plasmas in the absence of external torque.
Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas
NASA Astrophysics Data System (ADS)
Ribeiro, Celso
2015-11-01
The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.
Hu, J S; Sun, Z; Guo, H Y; Li, J G; Wan, B N; Wang, H Q; Ding, S Y; Xu, G S; Liang, Y F; Mansfield, D K; Maingi, R; Zou, X L; Wang, L; Ren, J; Zuo, G Z; Zhang, L; Duan, Y M; Shi, T H; Hu, L Q
2015-02-01
A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development. PMID:25699449
Plasma Profile and Shape Optimization for the Advanced Tokamak Power Plant, ARIES-AT
C.E. Kessel; T.K. Mau; S.C. Jardin; and F. Najmabadi
2001-06-05
An advanced tokamak plasma configuration is developed based on equilibrium, ideal-MHD stability, bootstrap current analysis, vertical stability and control, and poloidal-field coil analysis. The plasma boundaries used in the analysis are forced to coincide with the 99% flux surface from the free-boundary equilibrium. Using an accurate bootstrap current model and external current-drive profiles from ray-tracing calculations in combination with optimized pressure profiles, beta(subscript N) values above 7.0 have been obtained. The minimum current drive requirement is found to lie at a lower beta(subscript N) of 5.4. The external kink mode is stabilized by a tungsten shell located at 0.33 times the minor radius and a feedback system. Plasma shape optimization has led to an elongation of 2.2 and triangularity of 0.9 at the separatrix. Vertical stability could be achieved by a combination of tungsten shells located at 0.33 times the minor radius and feedback control coils located behind the shield. The poloidal-field coils were optimized in location and current, providing a maximum coil current of 8.6 MA. These developments have led to a simultaneous reduction in the power plant major radius and toroidal field.
Kenneth M. Young
2010-02-22
A Demonstration tokamak (Demo) is an essential next step toward a magnetic-fusion based reactor. One based on advanced-tokamak (AT) plasmas is especially appealing because of its relative compactness. However, it will require many plasma measurements to provide the necessary signals to feed to ancillary systems to protect the device and control the plasma. This note addresses the question of how much intrusion into the blanket system will be required to allow the measurements needed to provide the information required for plasma control. All diagnostics will require, at least, the same shielding designs as planned for ITER, while having the capability to maintain their calibration through very long pulses. Much work is required to define better the measurement needs and the quantity and quality of the measurements that will have to be made, and how they can be integrated into the other tokamak structures.
MHD stability of tokamak plasmas
Chance, M.S. Sun, Y.C.; Jardin, S.C.; Kessel, C.E.; Okabayashi, M.
1992-08-01
This paper will give an overview of the some of the methods which are used to simulate the ideal MHD properties of tokamak plasmas. A great deal of the research in this field is necessarily numerical and the substantial progress made during the past several years has roughly paralleled the continuing availability of more advanced supercomputers. These have become essential to accurately model the complex configurations necessary for achieving MHD stable reactor grade conditions. Appropriate tokamak MHD equilibria will be described. Then the stability properties is discussed in some detail, emphasizing the difficulties of obtaining stable high {beta} discharges in plasmas in which the current is mainly ohmically driven and thus demonstrating the need for tailoring the current and pressure profiles of the plasma away from the ohmic state. The outline of this paper will roughly follow the physics development to attain the second region of stability in the PBX-M device at The Princeton Plasmas Physics Laboratory.
Ding, B. J.; Kong, E. H.; Li, M. H.; Zhang, L.; Wei, W.; Li, Y. C.; Wu, J. H.; Xu, G. S.; Wang, M.; Gong, X. Z.; Shan, J. F.; Liu, F. K.; Zhang, T.; Ekedahl, A.; Zhao, H. L.; Collaboration: EAST Team
2013-10-15
Effect of gas puffing from electron-side and ion-side on lower hybrid wave (LHW)-plasma is investigated in experimental advanced superconductive tokamak for the first time. Experimental results with different gas flow rates show that electron density at the grill is higher in the case of gas puffing from electron-side; consequently, a lower reflection coefficient is observed, suggesting better effect of puffing from electron-side on LHW-plasma. The difference in edge density between electron- and ion-side cases suggests that local ionization of puffed gas plays a dominant role in affecting the density at the grill due to different movement direction of ionized electrons and that part of gas has been locally ionized near the gas pipe before diffusing into the grill region. Such difference could be enlarged and important in ITER due to the improvement of plasma parameters and LHW power.
Polarization spectroscopy of tokamak plasmas
Wroblewski, D.
1991-09-01
Measurements of polarization of spectral lines emitted by tokamak plasmas provide information about the plasma internal magnetic field and the current density profile. The methods of polarization spectroscopy, as applied to the tokamak diagnostic, are reviewed with emphasis on the polarimetry of motional Stark effect in hydrogenic neutral beam emissions. 25 refs., 7 figs.
NASA Astrophysics Data System (ADS)
Moreau, D.; Walker, M. L.; Ferron, J. R.; Liu, F.; Schuster, E.; Barton, J. E.; Boyer, M. D.; Burrell, K. H.; Flanagan, S. M.; Gohil, P.; Groebner, R. J.; Holcomb, C. T.; Humphreys, D. A.; Hyatt, A. W.; Johnson, R. D.; La Haye, R. J.; Lohr, J.; Luce, T. C.; Park, J. M.; Penaflor, B. G.; Shi, W.; Turco, F.; Wehner, W.; the ITPA-IOS Group members; experts
2013-06-01
The first real-time profile control experiments integrating magnetic and kinetic variables were performed on DIII-D in view of regulating and extrapolating advanced tokamak scenarios to steady-state devices and burning plasma experiments. Device-specific, control-oriented models were obtained from experimental data using a generic two-time-scale method that was validated on JET, JT-60U and DIII-D under the framework of the International Tokamak Physics Activity for Integrated Operation Scenarios (Moreau et al 2011 Nucl. Fusion 51 063009). On DIII-D, these data-driven models were used to synthesize integrated magnetic and kinetic profile controllers. The neutral beam injection (NBI), electron cyclotron current drive (ECCD) systems and ohmic coil provided the heating and current drive (H&CD) sources. The first control actuator was the plasma surface loop voltage (i.e. the ohmic coil), and the available beamlines and gyrotrons were grouped to form five additional H&CD actuators: co-current on-axis NBI, co-current off-axis NBI, counter-current NBI, balanced NBI and total ECCD power from all gyrotrons (with off-axis current deposition). Successful closed-loop experiments showing the control of (a) the poloidal flux profile, Ψ(x), (b) the poloidal flux profile together with the normalized pressure parameter, βN, and (c) the inverse of the safety factor profile, \\bar{\\iota}(x)=1/q(x) , are described.
Yang, Q. Q. Zhong, F. C. E-mail: fczhong@dhu.edu.cn; Jia, M. N.; Xu, G. S. E-mail: fczhong@dhu.edu.cn; Wang, L.; Wang, H. Q.; Chen, R.; Yan, N.; Liu, S. C.; Chen, L.; Li, Y. L.; Liu, J. B.
2015-06-15
The power fall-off width in the H-mode scrape-off layer (SOL) in tokamaks shows a strong inverse dependence on the plasma current, which was noticed by both previous multi-machine scaling work [T. Eich et al., Nucl. Fusion 53, 093031 (2013)] and more recent work [L. Wang et al., Nucl. Fusion 54, 114002 (2014)] on the Experimental Advanced Superconducting Tokamak. To understand the underlying physics, probe measurements of three H-mode discharges with different plasma currents have been studied in this work. The results suggest that a higher plasma current is accompanied by a stronger E×B shear and a shorter radial correlation length of turbulence in the SOL, thus resulting in a narrower power fall-off width. A simple model has also been applied to demonstrate the suppression effect of E×B shear on turbulence in the SOL and shows relatively good agreement with the experimental observations.
Kondoh, Yoshiomi; Fukasawa, Toshinobu
2009-11-15
Generalized simultaneous eigenvalue equations derived from a generalized theory of self-organization are applied to a set of simultaneous equations for two-fluid model plasmas. An advanced active control by using theoretical time constants is proposed by predicting quantities to be controlled. Typical high beta numerical configurations are presented for the ultra low q tokamak plasmas and the reversed-field pinch (RFP) ones in cylindrical geometry by solving the set of simultaneous eigenvalue equations. Improved confinement with no detectable saw-teeth oscillations in tokamak experiments is reasonably explained by the shortest time constant of ion flow. The shortest time constant of poloidal ion flow is shown to be a reasonable mechanism for suppression of magnetic fluctuations by pulsed poloidal current drives in RFP experiments. The bifurcation from basic eigenmodes to mixed ones deduced from stability conditions for eigenvalues is shown to be a good candidate for the experimental bifurcation from standard RFP plasmas to their improved confinement regimes.
Hu Chundong; Xie Yahong; Liu Sheng; Xie Yuanlai; Jiang Caichao; Song Shihua; Li Jun; Liu Zhimin
2011-02-15
High current ion source is the key part of the neutral beam injector. In order to develop the project of 4 MW neutral beam injection for the experimental advanced superconducting tokamak (EAST) on schedule, the megawatt high current ion source is prestudied in the Institute of Plasma Physics in China. In this paper, the megawatt high current ion source test bed and the first plasma are presented. The high current discharge of 900 A at 2 s and long pulse discharge of 5 s at 680 A are achieved. The arc discharge characteristic of high current ion source is analyzed primarily.
Advanced tokamak research on the DIII-D tokamak
Chan, V.S.
1994-01-01
The objective of the planned research in advanced tokamak development on DIII-D at General Atomics, San Diego, USA. is to establish improved tokamak operation through significant improvements in the stability factor, confinement quality, and bootstrap current fraction using localized radio frequency (rf) current profile control, rf and neutral beam heating for pressure profile control, as well as control of plasma rotation and optimization of the plasma boundary conditions. Recent research results in H-mode confinement, modifications of current profiles to achieve higher confinement and higher {beta}, a new regime of improved confinement (VH-mode), and rf noninductive current drive are encouraging. In this talk, arguments will be presented supporting the need for improved performance in tokamak reactors. Experimentally observed advanced performance regimes on DIII-D will be discussed. Confinement improvement up to H = 4, where H is the ratio of energy confinement time to the ITER89-P scaling H{triple_bond} {Tau}{sub E}/{Tau}{sub E-ITER89-P}, has been achieved. In other discharges {beta}{sub N} = {beta}/(I/aB),[%-m{center_dot}{Tau}/MA] {approx_gt} 6 has been obtained. These values have so far been achieved transiently and independently. Techniques, will be described which can extend the high performance to quasi-steady-state and sustain the high H and {beta}{sub N} values simultaneously. Two high performance regimes, one in first stable regime and the other in second stable regime, have been simulated br self-consistently evolving a magnetohydrodynamic (MHD) equilibrium-transport code. Finally, experimental program plans and outstanding important physics issues will be discussed.
Transport equations in tokamak plasmas
Callen, J. D.; Hegna, C. C.; Cole, A. J.
2010-05-15
Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for neoclassical effects on the parallel Ohm's law, fluctuation-induced transport, heating, current-drive and flow sources and sinks, small magnetic field nonaxisymmetries, magnetic field transients, etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed using a kinetic-based approach. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales and constraints they impose are considered sequentially: compressional Alfven waves (Grad-Shafranov equilibrium, ion radial force balance), sound waves (pressure constant along field lines, incompressible flows within a flux surface), and collisions (electrons, parallel Ohm's law; ions, damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on a plasma species: seven ambipolar collision-based ones (classical, neoclassical, etc.) and eight nonambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients, etc.). The plasma toroidal rotation equation results from setting to zero the net radial current induced by the nonambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the nonambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The 'mean field' effects of microturbulence on the parallel Ohm's law, poloidal ion flow, particle fluxes, and toroidal momentum and energy transport are all included self-consistently. The
ADX - Advanced Divertor and RF Tokamak Experiment
NASA Astrophysics Data System (ADS)
Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl
2015-11-01
The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.
Spherical tokamaks with plasma centre-post
NASA Astrophysics Data System (ADS)
Ribeiro, Celso
2013-10-01
The metal centre-post (MCP) in tokamaks is a structure which carries the total toroidal field current and also houses the Ohmic heating solenoid in conventional or low aspect ratio (Spherical)(ST) tokamaks. The MCP and solenoid are critical components for producing the toroidal field and for the limited Ohmic flux in STs. Constraints for a ST reactor related to these limitations lead to a minimum plasma aspect ratio of 1.4 which reduces the benefit of operation at higher betas in a more compact ST reactor. Replacing the MCP is of great interest for reactor-based ST studies since the device is simplified, compactness increased, and maintenance reduced. An experiment to show the feasibility of using a plasma centre-post (PCP) is being currently under construction and involves a high level of complexity. A preliminary study of a very simple PCP, which is ECR(Electron Cyclotron Resonance)-assisted and which includes an innovative fuelling system based on pellet injection, has recently been reported. This is highly suitable for an ultra-low aspect ratio tokamak (ULART) device. Advances on this PCP ECR-assisted concept within a ULART and the associated fuelling system are presented here, and will include the field topology for the PCP ECR-assisted scheme, pellet ablation modeling, and a possible global equilibrium simulation. VIE-ITCR, IAEA-CRP contr.17592, National Instruments-Costa Rica.
NASA Astrophysics Data System (ADS)
Hussain, Azam; Zhao, Zhenling; Xie, Jinlin; Zhu, Ping; Liu, Wandong; Ti, Ang
2016-04-01
The spatial and temporal evolutions of compound sawteeth were directly observed using 2D electron cyclotron emission imaging on experimental advanced superconducting tokamak. The compound sawtooth consists of partial and full collapses. After partial collapse, the hot core survives as only a small amount of heat disperses outwards, whereas in the following full collapse a large amount of heat is released and the hot core dissipates. The presence of two q = 1 surfaces was not observed. Instead, the compound sawtooth occurs mainly at the beginning of an ion cyclotron resonant frequency heating pulse and during the L-H transition phase, which may be related to heat transport suppression caused by a decrease in electron heat diffusivity.
OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS
LIN-LIU,YR; STAMBAUGH,RD
2002-11-01
OAK A271 OPTIMUM PLASMA STATES FOR NEXT STEP TOKAMAKS. The dependence of the ideal ballooning {beta} limit on aspect ratio, A, and elongation {kappa} is systematically explored for nearly 100% bootstrap current driven tokamak equilibria in a wide range of the shape parameters (A = 1.2-7.0, {kappa} = 1.5-6.0 with triangularity {delta} = 0.5). The critical {beta}{sub N} is shown to be optimal at {kappa} = 3.0-4.0 for all A studied and increases as A decreases with a dependence close to A{sup -0.5}. The results obtained can be used as a theoretical basis for the choice of optimum aspect ratio and elongation of next step burning plasma tokamaks or tokamak reactors.
Boundary Plasma Turbulence Simulations for Tokamaks
Xu, X; Umansky, M; Dudson, B; Snyder, P
2008-05-15
The boundary plasma turbulence code BOUT models tokamak boundary-plasma turbulence in a realistic divertor geometry using modified Braginskii equations for plasma vorticity, density (ni), electron and ion temperature (T{sub e}; T{sub i}) and parallel momenta. The BOUT code solves for the plasma fluid equations in a three dimensional (3D) toroidal segment (or a toroidal wedge), including the region somewhat inside the separatrix and extending into the scrape-off layer; the private flux region is also included. In this paper, a description is given of the sophisticated physical models, innovative numerical algorithms, and modern software design used to simulate edge-plasmas in magnetic fusion energy devices. The BOUT code's unique capabilities and functionality are exemplified via simulations of the impact of plasma density on tokamak edge turbulence and blob dynamics.
NASA Astrophysics Data System (ADS)
Yamazaki, K.; Uemura, S.; Oishi, T.; Garcia, J.; Arimoto, H.; Shoji, T.
2009-05-01
Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.
NASA Astrophysics Data System (ADS)
Xu, Y.; Xie, C. Y.; Qin, S. G.; Song, J. P.; Li, Q.; Zhao, S. X.; Liu, G. H.; Wang, T. J.; Yu, Y.; Luo, G.-N.
2014-04-01
To upgrade the Experimental Advanced Superconducting Tokamak dome and first-wall, flat-type W/Cu plasma-facing components will be installed in the coming years in order to exhaust the increasing heat flux. Mock-ups with an interlayer of oxygen-free Cu (OFC) made by vacuum hot pressing have been developed and the bonding strength was found to be over 100 MPa. The behavior of the mock-ups under steady-state high heat flux loads has been studied. No crack or exfoliation occurred on the W surface and W/OFC/CuCrZr interfaces after screening tests with heat fluxes of 2.24-7.73 MW m-2. The mock-up survived up to 1000 cycles heat load of 3.24 MW m-2 with cooling water of 4 m s-1, 20 °C. However, cracks appeared in W around the gaps at about the 300th cycle under a heat load of 5.37 MW m-2. We have also studied the chemical vapor deposition W coated CuCrZr with an OFC interlayer. It has been found that: (i) the OFC interlayer plays a significant role in achieving coatings without any crack, (ii) the deposition rate was about 0.3-0.5 mm h-1 at 490-580 °C and (iii) a bonding strength of 53.7 MPa was achieved with laser surfi-sculpt.
LIDAR Thomson scattering for advanced tokamaks. Final report
Molvik, A.W.; Lerche, R.A.; Nilson, D.G.
1996-03-18
The LIDAR Thomson Scattering for Advanced Tokamaks project made a valuable contribution by combining LLNL expertise from the MFE Program: tokamak design and diagnostics, and the ICF Program and Physics Dept.: short-pulse lasers and fast streak cameras. This multidisciplinary group evaluated issues involved in achieving a factor of 20 higher high spatial resolution (to as small as 2-3 mm) from the present state of the art in LIDAR Thomson scattering, and developed conceptual designs to apply LIDAR Thomson scattering to three tokamaks: Upgraded divertor measurements in the existing DIII-D tokamak; Both core and divertor LIDAR Thomson scattering in the proposed (now cancelled) TPX; and core, edge, and divertor LIDAR Thomson scattering on the presently planned International Tokamak Experimental Reactor, ITER. Other issues were evaluated in addition to the time response required for a few millimeter spatial resolution. These include the optimum wavelength, 100 Hz operation of the laser and detectors, minimizing stray light - always the Achilles heel of Thomson scattering, and time dispersion in optics that could prevent good spatial resolution. Innovative features of our work included: custom short pulsed laser concepts to meet specific requirements, use of a prism spectrometer to maintain a constant optical path length for high temporal and spatial resolution, the concept of a laser focus outside the plasma to ionize gas and form an external fiducial to use in locating the plasma edge as well as to spread the laser energy over a large enough area of the inner wall to avoid laser ablation of wall material, an improved concept for cleaning windows between shots by means of laser ablation, and the identification of a new physics issue - nonlinear effects near a laser focus which could perturb the plasma density and temperature that are to be measured.
NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK
WALKER, ML; FERRON, JR; HUMPHREYS, DA; JOHNSON, RD; LEUER, JA; PENAFLOR, BG; PIGLOWSKI, DA; ARIOLA, M; PIRONTI, A; SCHUSTER, E
2002-10-01
OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability.
Electrostatic analysis of the tokamak edge plasma
Motley, R.W.
1981-07-01
The intrusion of an equipotential poloidal limiter into the edge plasma of a circular tokamak discharge distorts the axisymmetry in two ways: (1) it (partially) shorts out the top-to-bottom Pfirsch-Schlueter driving potentials, and (2) it creates zones of back current flow into the limiter. The resulting boundary mismatch between the outer layers and the inner axisymmetric Pfirsch-Schlueter layer provides free energy to drive the edge plasma unstable. Special limiters are proposed to symmetrize the edge plasma and thereby reduce the electrical and MHD activity in the boundary layer.
Forced Magnetic Reconnection In A Tokamak Plasma
NASA Astrophysics Data System (ADS)
Callen, J. D.; Hegna, C. C.
2015-11-01
The theory of forced magnetic field reconnection induced by an externally imposed resonant magnetic perturbation usually uses a sheared slab or cylindrical magnetic field model and often focuses on the potential time-asymptotic induced magnetic island state. However, tokamak plasmas have significant magnetic geometry and dynamical plasma toroidal rotation screening effects. Also, finite ion Larmor radius (FLR) and banana width (FBW) effects can damp and thus limit the width of a nascent magnetic island. A theory that is more applicable for tokamak plasmas is being developed. This new model of the dynamics of forced magnetic reconnection considers a single helicity magnetic perturbation in the tokamak magnetic field geometry, uses a kinetically-derived collisional parallel electron flow response, and employs a comprehensive dynamical equation for the plasma toroidal rotation frequency. It is being used to explore the dynamics of bifurcation into a magnetically reconnected state in the thin singular layer around the rational surface, evolution into a generalized Rutherford regime where the island width exceeds the singular layer width, and assess the island width limiting effects of FLR and FBW polarization currents. Support by DoE grants DE-FG02-86ER53218, DE-FG02-92ER54139.
LONG PULSE ADVANCED TOKAMAK DISCHARGES IN THE DIII-D TOKAMAK
P.I. PETERSEN
2002-06-01
One of the main goals for the DIII-D research program is to establish an advanced tokamak plasma with high bootstrap current fraction that can be sustained in-principle steady-state. Substantial progress has been made in several areas during the last year. The resistive wall mode stabilization has been done with spinning plasmas in which the plasma pressure has been extended well above the no-wall beta limit. The 3/2 neoclassical tearing mode has been stabilized by the injection of ECH into the magnetic islands, which drives current to substitute the missing bootstrap current. In these experiments either the plasma was moved or the toroidal field was changed to overlap the ECCD resonance with the location of the NTMs. Effective disruption mitigation has been obtained by massive noble gas injection into shots where disruptions were deliberately triggered. The massive gas puff causes a fast and clean current quench with essentially all the plasma energy radiated fairly uniformly to the vessel walls. The run-away electrons that are normally seen accompanying disruptions are suppressed by the large density of electrons still bound on the impurity nuclei. Major elements required to establish integrated, long-pulse, advanced tokamak operations have been achieved in DIII-D: {beta}{sub T} = 4.2%, {beta}{sub p} = 2, f{sub BS} = 65%, and {beta}{sub N}H{sub 89} = 10 for 600 ms ({approx} 4{tau}{sub E}). The next challenge is to integrate the different elements, which will be the goal for the next five years when additional control will be available. Twelve resistive wall mode coils are scheduled to be installed in DIII-D during the summer of 2003. The future plans include upgrading the tokamak pulse length capability and increasing the ECH power, to control the current profile evolution.
High beta plasmas in the PBX tokamak
Bol, K.; Buchenauer, D.; Chance, M.; Couture, P.; Fishman, H.; Fonck, R.; Gammel, G.; Grek, B.; Ida, K.; Itami, K.
1986-04-01
Bean-shaped configurations favorable for high ..beta.. 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 ..beta.. limit.
Plasma Physics Regimes in Tokamaks with Li Walls
L.E. Zakharo; N.N. Gorelenkov; R.B. White; S.I. Krasheninnikov; G.V. Pereverzev
2003-08-21
Low recycling regimes with a plasma limited by a lithium wall surface suggest enhanced stability and energy confinement, both necessary for tokamak reactors. These regimes could make ignition feasible in compact tokamaks. Ignited Spherical Tokamaks (IST), self-sufficient in the bootstrap current, are introduced as a necessary step for development of the physics and technology of power reactors.
Spontaneous generation of rotation in tokamak plasmas
Parra Diaz, Felix
2013-12-24
Three different aspects of intrinsic rotation have been treated. i) A new, first principles model for intrinsic rotation [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has been implemented in the gyrokinetic code GS2. The results obtained with the code are consistent with several experimental observations, namely the rotation peaking observed after an L-H transition, the rotation reversal observed in Ohmic plasmas, and the change in rotation that follows Lower Hybrid wave injection. ii) The model in [F.I. Parra, M. Barnes and P.J. Catto, Nucl. Fusion 51, 113001 (2011)] has several simplifying assumptions that seem to be satisfied in most tokamaks. To check the importance of these hypotheses, first principles equations that do not rely on these simplifying assumptions have been derived, and a version of these new equations has been implemented in GS2 as well. iii) A tokamak cross-section that drives large intrinsic rotation has been proposed for future large tokamaks. In large tokamaks, intrinsic rotation is expected to be very small unless some up-down asymmetry is introduced. The research conducted under this contract indicates that tilted ellipticity is the most efficient way to drive intrinsic rotation.
Numerical investigations of plasma parameters in the COMPASS tokamak
Havlickova, E.; Zagorski, R.; Panek, R.
2008-09-15
A numerical investigation of plasma parameters in a diverter configuration of COMPASS tokamak is presented. The plasma parameters in the device are analyzed in the frame of the self-consistent description of the central plasma and edge region. The possibility of achieving high recycling and detached regimes in the boundary layer of the COMPASS tokamak is discussed.
MHD Effects of a Ferritic Wall on Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Hughes, Paul E.
It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency
Theory of "clumps" in drift-wave turbulence in tokamak plasma
NASA Astrophysics Data System (ADS)
Wang, Xiaogang; Qiu, Xiaoming; X, M. Qhiu
1986-08-01
Basing on the new method of trajectory stochastic treatment advanced by one of the authors of this paper, the theory of "clumps" in driftwave turbulence in tokamak plasmas has been developed. It is shown that, as a longer time behaviour, plasmas in tokamaks will have the same "clumps" effects as those in uniform magnetic fields, although the diffusion crossing magnetic field lines in tokamaks will be enhanced. The influence of the non-uniformity of the magnetic field, such as curvature, shear, etc., on the transverse diffusion and the "clump" life-time is discussed.
Argonne Plasma Engineering Experiment (APEX) Tokamak
Norem, J.H.; Balka, L.J.; Kulovitz, E.E.; Magill, S.R.; McGhee, D.G.; Moretti, A.; Praeg, W.F.
1981-03-01
The Argonne Plasma Engineering Experiment (APEX) Tokamak was designed to provide hot plasmas for reactor-relevant experiments with rf heating (current drive) and plasma wall experiments, principally in-situ low-Z wall coating and maintenance. The device, sized to produce energetic plasmas at minimum cost, is small (R = 51 cm, r = 15 cm) but capable of high currents (100 kA) and long pulse durations (100 ms). A design using an iron central core with no return legs, pure tension tapewound toroidal field coils, digital radial position control, and UHV vacuum technology was used. Diagnostics include monochrometers, x-ray detectors, and a microwave interferometer and radiometer for density and temperature measurements. Stable 100 ms shots were produced with electron temperatures in the range 500 to 1000 eV. Initial results included studies of thermal desorption and recoating of wall materials.
Plasma transport in a compact ignition tokamak
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 /sup 3/He minority cyclotron resonance layer. Under these conditions, sawtooth oscillations are small and have little impact on ignition. Tritium inventory is minimized by preconditioning most discharges with deuterium. Tritium is injected, in large frozen pellets, only after minority resonance preheating. Variations of the transport model, impurity influx, heating profile, and pellet ablation rates, have a large effect on ignition and on the maximum beta that can be achieved.
Mathematical modeling plasma transport in tokamaks
Quiang, Ji
1995-12-31
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{sup 20}/m{sup 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%.
Development of a tokamak plasma optimized for stability and confinement
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 ({beta}) and high energy confinement ({tau}{sub E}); either of these, by itself, is insufficient. In operation of the DIII-D tokamak, both high confinement enhancement (H{equivalent_to} {tau}{sub E}/{tau}{sub ITER-89P} = 4) and high normalized {beta} ({beta}{sub N}{equivalent_to} {beta}/(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{sub min} = 2.6, and q{sub 95} = 6]. This model plasma uses profiles which the authors expect to be realizable. At {beta}{sub 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.
Neoclassical Transport Properties of Tokamak Plasmas
Weyssow, B.
2004-03-15
The classical transport theory is strictly valid for a plasma in a homogeneous and stationary magnetic field. In the '60, experiments have shown that this theory does not apply as a local theory of transport in Tokamaks. It was shown that global geometric characteristics of the confining elements have a strong influence on the transport. Three regimes of collisionality are characteristic of the neoclassical transport theory: the banana regime (the electronic diffusion coefficient increases starting from zero), the plateau regime (the diffusion coefficient is almost independent of the collisionality) and the Pfirsch-Schlueter regime (the electronic diffusion coefficient again increases with the collisionality)
Viscosity in the edge of tokamak plasmas
NASA Astrophysics Data System (ADS)
Stacey, W. M.
1993-05-01
A fluid representation of viscosity has been incorporated into a set of fluid equations that are maximally ordered in the 'short radial gradient scale length' (srgsl) ordering that is appropriate for the edge of tokamak plasmas. The srgsl ordering raises viscous drifts and other viscous terms to leading order and fundamentally alters the character of the fluid equations. A leasing order viscous drift is identified. Viscous-driven radial particle and energy fluxes in the scrape-off layer and divertor channel are estimated to have an order unity effect in reducing radial peaking of energy fluxes transported along the field lines to divertor collector plates.
Neoclassical diffusion of heavy impurities in a rotating tokamak plasma
Wong, K.L.; Cheng, C.Z.
1987-08-01
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, resulting in a diffusion coefficient much larger than that for a stationary plasma. Particle simulation is carried out and the results offer a qualitative explanation for some experimental data from the Tokamak Test Reactor (TFTR). 13 refs., 2 figs.
Modeling of ICRF Internal Transport Barrier Control for Advanced Tokamaks
NASA Astrophysics Data System (ADS)
Sund, R. S.; Scharer, J. E.
1998-11-01
We present an analysis of TFTR ICRF current drive experiments carried out by Majeski et al.(R. Majeski, J. Rodgers, G. Schilling, C. Phillips, J. Hosea and the TFTR Group, private communication.) The influence of deuterium, tritium, minority specie, electron and alpha concentrations, temperatures and beam fractions are considered for the two-ion mode conversion current drive experiments. Direct comparison with experimental data is carried out by means of a nonlocal large gyroradius ICRF code(O. Sauter, Ph.D. thesis, Ecole Polytechnique de Lausanne, Switzerland (1992).) which incorporates 1-D plasma profiles. It is found that substantial beam and alpha particle absorption can occur for some cases. Application of ion cyclotron range of frequencies internal transport barrier control requires further examination of fast wave mode conversion and the interaction of ion Bernstein waves with plasmas in advanced tokamaks. The effects of perpendicular and parallel magnetic gradients on the ion, electron, and alpha particle absorption are examined. A viable internal transport barrier control scheme for a reactor grade advanced tokamak will be discussed.
Nonlinear lower hybrid modeling in tokamak plasmas
Napoli, F.; Schettini, G.; Castaldo, C.; Cesario, R.
2014-02-12
We present here new results concerning the nonlinear mechanism underlying the observed spectral broadening produced by parametric instabilities occurring at the edge of tokamak plasmas in present day LHCD (lower hybrid current drive) experiments. Low frequency (LF) ion-sound evanescent modes (quasi-modes) are the main parametric decay channel which drives a nonlinear mode coupling of lower hybrid (LH) waves. The spectrum of the LF fluctuations is calculated here considering the beating of the launched LH wave at the radiofrequency (RF) operating line frequency (pump wave) with the noisy background of the RF power generator. This spectrum is calculated in the frame of the kinetic theory, following a perturbative approach. Numerical solutions of the nonlinear LH wave equation show the evolution of the nonlinear mode coupling in condition of a finite depletion of the pump power. The role of the presence of heavy ions in a Deuterium plasma in mitigating the nonlinear effects is analyzed.
Stability of tearing modes in tokamak plasmas
Hegna, C.C.; Callen, J.D.
1994-02-01
The stability properties of m {ge} 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 {Delta}{prime}. This calculation generalizes previous work on this topic by considering more general toroidal equilibria (however, toroidal coupling effects are ignored). Constructions of {Delta}{prime} 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.
Plasma diagnostics for the compact ignition tokamak
Medley, S.S.; Young, K.M.
1988-06-01
The primary mission of the Compact Ignition Tokamak (CIT) is to study the physics of alpha-particle heating in an ignited D-T plasma. A burn time of about 10 /tau//sub E/ is projected in a divertor configuration with baseline machine design parameters of R=2.10 m, 1=0.65 m, b=1.30 m, I/sub p/=11 MA, B/sub T/=10 T and 10-20 MW of auxiliary rf heating. Plasma temperatures and density are expected to reach T/sub e/(O) /approximately/20 keV, T/sub i/(O) /approximately/30 keV, and n/sub e/(O) /approximately/ 1 /times/ 10/sup 21/m/sup /minus/3/. The combined effects of restricted port access to the plasma, the presence of severe neutron and gamma radiation backgrounds, and the necessity for remote of in-cell components create challenging design problems for all of the conventional diagnostic associated with tokamak operations. In addition, new techniques must be developed to diagnose the evolution in space, time, and energy of the confined alpha distribution as well as potential plasma instabilities driven by collective alpha-particle effects. The design effort for CIT diagnostics is presently in the conceptual phase with activity being focused on the selection of a viable diagnostic set and the identification of essential research and development projects to support this process. A review of these design issues and other aspects impacting the selection of diagnostic techniques for the CIT experiment will be presented. 28 refs., 10 figs., 2 tabs.
DIII-D Advanced Tokamak Research Overview
V.S. Chan; C.M. Greenfield; L.L. Lao; T.C. Luce; C.C. Petty; G.M. Staebler
1999-12-01
This paper reviews recent progress in the development of long-pulse, high performance discharges on the DIII-D tokamak. It is highlighted by a discharge achieving simultaneously {beta}{sub N}H of 9, bootstrap current fraction of 0.5, noninductive current fraction of 0.75, and sustained for 16 energy confinement times. The physics challenge has changed in the long-pulse regime. Non-ideal MHD modes are limiting the stability, fast ion driven modes may play a role in fast ion transport which limits the stored energy and plasma edge behavior can affect the global performance. New control tools are being developed to address these issues.
Advanced ICRF antenna design for R-TOKAMAK
NASA Astrophysics Data System (ADS)
Kako, E.; Ando, R.; Ichimura, M.; Ogawa, Y.; Amano, T.; Watari, T.
1986-01-01
The advanced ICRF antennas designed for the R-TOKAMAK (a proposal in the Institute of Plasma Physics, Nagoya University) are described. They are a standard loop antenna and a panel heater antenna for fast wave heating, and a waveguide antenna for ion Bernstein wave heating. The standard loop antenna is made of Al-alloy and has a simple structure to install because of radioactivation by D-T neutrons. For high power heating, a new type antenna called Panel heater antenna is proposed. It has a wide radiation area and is able to select a parallel wave number k. The field pattern of the panel heater antenna is measured. The feasibility of the waveguide antenna is discussed for ion Bernstein wave heating. The radiation from the aperture of the double ridge waveguide is experimentally estimated with a load simulating the plasma.
Multiscale coherent structures in tokamak plasma turbulence
Xu, G. S.; Wan, B. N.; Zhang, W.; Yang, Q. W.; Wang, L.; Wen, Y. Z.
2006-10-15
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.
Plasma Confinement in the UCLA Electric Tokamak.
NASA Astrophysics Data System (ADS)
Taylor, Robert J.
2001-10-01
The main goal of the newly constructed large Electric Tokamak (R = 5 m, a = 1 m, BT < 0.25 T) is to access an omnigeneous, unity beta(S.C. Cowley, P.K. Kaw, R.S. Kelly, R.M. Kulsrud, Phys. fluids B 3 (1991) 2066.) plasma regime. The design goal was to achieve good confinement at low magnetic fields, consistent with the high beta goal. To keep the program cost down, we adopted the use of ICRF as the primary heating source. Consequently, antenna surfaces covering 1/2 of the surface of the tokamak has been prepared for heating and current drive. Very clean hydrogenic plasmas have been achieved with loop voltage below 0.7 volt and densities 3 times above the Murakami limit, n(0) > 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
Plasma/Liquid-Metal Interactions During Tokamak Operation
Hassanein, A.; Allain, J.P.; Insepov, Z.; Konkashbaev, I.
2005-04-15
One of the critical technological challenges of future tokamak fusion devices is the ability for plasma-facing components to handle both normal and abnormal plasma/surface interaction events that compromise their lifetime and operation of the machine. Under normal operation plasma/surface interactions that are important include: sputtering, particle implantation and recycling, He pumping and ELM (edge localized modes)-induced erosion. In abnormal or off-normal operation: disruptions and vertical displacement events (VDEs) are important. To extend PFC lifetime under these conditions, liquid-metals have been considered as candidate PFCs (Plasma-Facing Components), including: liquid lithium, tin-lithium, gallium and tin.Liquid lithium has been measured to have nonlinear increase of physical sputtering with rise in temperature. Such increase can be a result of exposure to ELM-level particle fluxes. The significant increase in particle flux to the divertor and nearby PFCs can enhance sputtering erosion by an order of magnitude or more. In addition from the standpoint of hydrogen recycling and helium pumping liquid lithium appears to be a good candidate plasma-facing material (PFM). Advanced designs of first wall and divertor systems propose the application of liquid-metals as an alternate PFC to contend with high-heat flux constraints of large-scale tokamak devices. Additional issues include PFC operation under disruptions and long temporal instabilities such as VDEs. A comprehensive two-fluid model is developed to integrate core and SOL (scrape-off layer) parameters during ELMs with PFC surface evolution using the HEIGHTS package. Special emphasis is made on the application of lithium as a candidate plasma-facing liquid-metal.
Plasma/liquid metal interactions during tokamak operation.
Hassanein, A.; Allain, J. P.; Insepov, Z.; Konkashbaev, I.; Energy Technology
2005-04-01
One of the critical technological challenges of future tokamak fusion devices is the ability for plasma-facing components to handle both normal and abnormal plasma/surface interaction events that compromise their lifetime and operation of the machine. Under normal operation plasma/surface interactions that are important include: sputtering, particle implantation and recycling, He pumping and ELM (edge localized modes)-induced erosion. In abnormal or off-normal operation: disruptions and vertical displacement events (VDEs) are important. To extend PFC lifetime under these conditions, liquid-metals have been considered as candidate PFCs (Plasma-Facing Components), including: liquid lithium, tin-lithium, gallium and tin. Liquid lithium has been measured to have nonlinear increase of physical sputtering with rise in temperature. Such increase can be a result of exposure to ELM-level particle fluxes. The significant increase in particle flux to the divertor and nearby PFCs can enhance sputtering erosion by an order of magnitude or more. In addition from the standpoint of hydrogen recycling and helium pumping liquid lithium appears to be a good candidate plasma-facing material (PFM). Advanced designs of first wall and divertor systems propose the application of liquid-metals as an alternate PFC to contend with high-heat flux constraints of large-scale tokamak devices. Additional issues include PFC operation under disruptions and long temporal instabilities such as VDEs. A comprehensive two-fluid model is developed to integrate core and SOL (scrape-off layer) parameters during ELMs with PFC surface evolution using the HEIGHTS package. Special emphasis is made on the application of lithium as a candidate plasma-facing liquid-metal.
Transport Bifurcation in a Rotating Tokamak Plasma
Highcock, E. G.; Barnes, M.; Schekochihin, A. A.; Parra, F. I.; Roach, C. M.; Cowley, S. C.
2010-11-19
The effect of flow shear on turbulent transport in tokamaks is studied numerically in the experimentally relevant limit of zero magnetic shear. It is found that the plasma is linearly stable for all nonzero flow shear values, but that subcritical turbulence can be sustained nonlinearly at a wide range of temperature gradients. Flow shear increases the nonlinear temperature gradient threshold for turbulence but also increases the sensitivity of the heat flux to changes in the temperature gradient, except over a small range near the threshold where the sensitivity is decreased. A bifurcation in the equilibrium gradients is found: for a given input of heat, it is possible, by varying the applied torque, to trigger a transition to significantly higher temperature and flow gradients.
TFTR/JET INTOR workshop on plasma transport tokamaks
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.
Seo, Seong-Heon; Wi, H. M.; Lee, W. R.; Kim, H. S.; Lee, T. G.; Kim, Y. S.; Park, Jinhyung; Kang, Jin-Seob; Bog, M. G.; Yokota, Y.; Mase, A.
2013-08-15
Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6–54 GHz), V band (48–72 GHz), and W band (72–108 GHz) to measure the density up to 7 × 10{sup 19} m{sup −3} when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.
The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study
Kessel, C. E.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Tillack, M. S.; Najmabadi, F.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; El-Guebaly, L.; Blanchard, J. P.; Martin, C. J.; Mynsburge, L.; Humrickhouse, P.; Rensink, M. E.; Rognlien, T. D.; Yoda, M.; Abdel-Khalik, S. I.; Hageman, M. D.; Mills, B. H.; Radar, J. D.; Sadowski, D. L.; Snyder, P. B.; St. John, H.; Turnbull, A. D.; Waganer, L. M.; Malang, S.; Rowcliffe, A. F.
2014-03-05
Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a {beta}N{sup total} of 5.75, H{sub 98} of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m{sup 2}. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a {beta}N{sup total} of 2.5, H{sub 98} of 1.25, n/n{sub Gr} of 1.3, and peak divertor heat flux of 10 MW/m{sup 2}. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m{sup 2}. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.
The ARIES Advanced and Conservative Tokamak Power Plant Study
Kessel, C. E; Tillak, M. S; Najmabadi, F.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; EL-Guebaly, L.; Blanchard, J. P.; Martin, C. J.; Mynsburge, L.; Humrickhouse, P.; Rensink, M. E.; Rognlien, T. D.; Yoda, M.; Abdel-Khalik, S. I.; Hageman, M. D.; Mills, B. H.; Rader, J. D.; Sadowski, D. L.; Snyder, P. B.; St. John, H.; Turnbull, A. D.; Waganer, L. M.; Malang, S.; Rowcliffe, A. F.
2015-12-22
Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦ^{total} _{N} of 5.75, an H98 of 1.65, an n/n_{Gr} of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦ^{total}_{N} of 2.5, an H₉₈ of 1.25, an n/n_{Gr} of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.
The ARIES Advanced and Conservative Tokamak Power Plant Study
Kessel, C. E; Tillak, M. S; Najmabadi, F.; Poli, F. M.; Ghantous, K.; Gorelenkov, N.; Wang, X. R.; Navaei, D.; Toudeshki, H. H.; Koehly, C.; et al
2015-12-22
Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦtotal N of 5.75, an H98 of 1.65, anmore » n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦtotalN of 2.5, an H₉₈ of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less
Orbit effects on impurity transport in a rotating tokamak plasma
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.
Ionization balance in EBIT and tokamak plasmas
NASA Astrophysics Data System (ADS)
Peacock, N. J.; Barnsley, R.; O'Mullane, M. G.; Tarbutt, M. R.; Crosby, D.; Silver, J. D.; Rainnie, J. A.
2001-01-01
The equilibrium state in tokamak core plasmas has been studied using the relative intensities of resonance x-ray lines, for example Lyα (H-like), "w" (He-like), and "q" (Li-like) from test ions such as Ar+15, Ar+16, and Ar+17. A full spatial analysis involves comparison of the line intensities with ion diffusion calculations, including relevant atomic rates. A zero-dimensional model using a global ion loss rate approximation has also been demonstrated by comparison with the data collected from a Johann configuration spectrometer with a charged coupled device (CCD) detector. Since the lines are nearly monoenergetic, their intensities are independent of the instrument sensitivity and are directly proportional to the ion abundances. This method has recently been applied to Ar in the Oxford electron beam ion trap (EBIT) with a beam energy in the range 3-10 keV. Taking into account the cross sections for monoenergetic electron collisions and polarization effects, model calculations agree with the observed line ratios at 4.1 keV beam energy. This work will be expanded to provide nomograms of ionization state versus line intensity ratios as a function of EBIT beam energy.
Designing a tokamak fusion reactor—How does plasma physics fit in?
NASA Astrophysics Data System (ADS)
Freidberg, J. P.; Mangiarotti, F. J.; Minervini, J.
2015-07-01
This paper attempts to bridge the gap between tokamak reactor design and plasma physics. The analysis demonstrates that the overall design of a tokamak fusion reactor is determined almost entirely by the constraints imposed by nuclear physics and fusion engineering. Virtually, no plasma physics is required to determine the main design parameters of a reactor: a , R 0 , B 0 , T i , T e , p , n , τ E , I . The one exception is the value of the toroidal current I , which depends upon a combination of engineering and plasma physics. This exception, however, ultimately has a major impact on the feasibility of an attractive tokamak reactor. The analysis shows that the engineering/nuclear physics design makes demands on the plasma physics that must be satisfied in order to generate power. These demands are substituted into the well-known operational constraints arising in tokamak physics: the Troyon limit, Greenwald limit, kink stability limit, and bootstrap fraction limit. Unfortunately, a tokamak reactor designed on the basis of standard engineering and nuclear physics constraints does not scale to a reactor. Too much current is required to achieve the necessary confinement time for ignition. The combination of achievable bootstrap current plus current drive is not sufficient to generate the current demanded by the engineering design. Several possible solutions are discussed in detail involving advances in plasma physics or engineering. The main contribution of the present work is to demonstrate that the basic reactor design and its plasma physics consequences can be determined simply and analytically. The analysis thus provides a crisp, compact, logical framework that will hopefully lead to improved physical intuition for connecting plasma physic to tokamak reactor design.
Recent results from tokamak divertor plasma measurements
Allen, S.L.
1996-05-01
New diagnostics have been developed to address key divertor physics questions, including: target plate heat flux reduction by radiation, basic edge transport issues, and plasma wall interactions (PWI) such as erosion. A system of diagnostics measures the target plate heat flux (imaging IR thermography) and particle flux (probes, pressure and Penning gauges, and visible emission arrays). Recently, T{sub e},n{sub e}, and P{sub e} (electron pressure) have been measured in 2-D with divertor Thomson Scattering. During radiative divertor operation T{sub e} is less than 2 eV, indicating that new atomic processes are important. Langmuir probes measure higher T{sub e} in some cases. In addition, the measured P{sub e} near the separatrix at the target plate is lower than the midplane pressure, implying radial momentum transport. Bolometer arrays, inverted with reconstruction algorithms, provide the 2-D core and divertor radiation profiles. Spectroscopic measurements identify the radiating species and provide information on impurity transport; both absolute chordal measurements and tomographic reconstructions of images are used. Either intrinsic carbon or an inert species (e.g., injected Ne) are usually observed, and absolute particle inventories are obtained. Computer codes are both benchmarked with the experimental data and provide important consistency checks. Several techniques are used to measure fundamental plasma transport and fluctuations, including probes and reflectometry. PWI issues are studied with in-situ coupons and insertable samples (DiMES). Representative divertor results from DIII-D with references to results on other tokamaks will be presented.
Hollow current profile scenarios for advanced tokamak reactor operations
Gourdain, P.-A.; Leboeuf, J.-N.
2009-11-15
Advanced tokamak scenarios are a possible approach to boosting reactor performances. Such schemes usually trigger current holes, a particular magnetohydrodynamics equilibrium where no current or pressure gradients exist in the core of the plasma. While such equilibria have large bootstrap fractions, flat pressure profiles in the plasma core may not be optimal for a reactor. However, moderate modifications of the equilibrium current profile can lead to diamagnetism where most of the pressure gradient is now balanced by poloidal currents and the toroidal magnetic field. In this paper, we consider the properties of diamagnetic current holes, also called ''dual equilibria,'' and demonstrate that fusion throughput can be significantly increased in such scenarios. Their stability is investigated using the DCON code. Plasmas with a beta peak of 30% and an average beta of 6% are found stable to both fixed and free-boundary modes with toroidal mode numbers n=1-4, as well as Mercier and high-n ballooning modes. This is not surprising as these scenarios have a normal beta close to 3.
Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak.
Qu, Hao; Zhang, Tao; Han, Xiang; Wen, Fei; Zhang, Shoubiao; Kong, Defeng; Wang, Yumin; Gao, Yu; Huang, Canbin; Cai, Jianqing; Gao, Xiang
2015-08-01
An X-mode polarized V band (50 GHz-75 GHz) radial and poloidal correlation reflectometry is designed and installed on Experimental Advanced Superconducting Tokamak (EAST). Two frequency synthesizers (12 GHz-19 GHz) are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together for launching through one single pyramidal antenna. Two poloidally separated antennae are installed to receive the reflected waves from plasma. This reflectometry system can be used for radial and poloidal correlation measurement of the electron density fluctuation. In ohmically heated plasma, the radial correlation length is about 1.5 cm measured by the system. The poloidal correlation analysis provides a means to estimate the fluctuation velocity perpendicular to the main magnetic field. In the present paper, the distance between two poloidal probing points is calculated with ray-tracing code and the propagation time is deduced from cross-phase spectrum. Fluctuation velocity perpendicular to the main magnetic field in the core of ohmically heated plasma is about from -1 km/s to -3 km/s. PMID:26329188
Radial and poloidal correlation reflectometry on Experimental Advanced Superconducting Tokamak
Qu, Hao; Zhang, Tao; Han, Xiang; Wen, Fei; Zhang, Shoubiao; Kong, Defeng; Wang, Yumin; Gao, Yu; Huang, Canbin; Cai, Jianqing; Gao, Xiang
2015-08-15
An X-mode polarized V band (50 GHz–75 GHz) radial and poloidal correlation reflectometry is designed and installed on Experimental Advanced Superconducting Tokamak (EAST). Two frequency synthesizers (12 GHz–19 GHz) are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together for launching through one single pyramidal antenna. Two poloidally separated antennae are installed to receive the reflected waves from plasma. This reflectometry system can be used for radial and poloidal correlation measurement of the electron density fluctuation. In ohmically heated plasma, the radial correlation length is about 1.5 cm measured by the system. The poloidal correlation analysis provides a means to estimate the fluctuation velocity perpendicular to the main magnetic field. In the present paper, the distance between two poloidal probing points is calculated with ray-tracing code and the propagation time is deduced from cross-phase spectrum. Fluctuation velocity perpendicular to the main magnetic field in the core of ohmically heated plasma is about from −1 km/s to −3 km/s.
A Midsize Tokamak As Fast Track To Burning Plasmas
E. Mazzucato
2010-07-14
This paper presents a midsize tokamak as a fast track to the investigation of burning plasmas. It is shown that it could reach large values of energy gain (≥10) with only a modest improvement in confinement over the scaling that was used for designing the International Thermonuclear Experimental Reactor (ITER). This could be achieved by operating in a low plasma recycling regime that experiments indicate can lead to improved plasma confinement. The possibility of reaching the necessary conditions of low recycling using a more efficient magnetic divertor than those of present tokamaks is discussed.
Development of a free-boundary tokamak equilibrium solver for advanced study of tokamak equilibria
NASA Astrophysics Data System (ADS)
Jeon, Young Mu
2015-09-01
A free-boundary Tokamak equilibrium solver (TES), developed for advanced study of tokamak equilibra, is described with two distinctive features. One is a generalized method to resolve the intrinsic axisymmetric instability, which is encountered in all equilibrium calculations with a freeboundary condition. The other is an extension to deal with a new divertor geometry such as snowflake or X divertors. For validations, the uniqueness of a solution is confirmed by the independence of variations in the computational domain, the mathematical correctness and accuracy of equilibrium profiles are checked by using a direct comparison with an analytic equilibrium known as a generalized Solov'ev equilibrium, and the governing force balance relation is tested by examining the intrinsic axisymmetric instabilities. As an application of an advanced equilibrium study, a snow-flake divertor configuration that requires a second-order zero of the poloidal magnetic flux is discussed in the circumstance of the Korea superconducting tokamak advanced research (KSTAR) coil system.
Waves and turbulence in a tokamak fusion plasma.
Surko, C M; Slusher, R E
1983-08-26
The tokamak is a prototype fusion device in which a toroidal Magnetic field is used to confine a hot plasma. Coherent waves, excited near the plasma edge, can be used to transport energy into the plasma in order to heat it to the temperatures required for thermonuclear fusion. In addition, tokamak plasmas are known to exhibit high levels of turbulent density fluctuations, which can transport particles and energy out of the plasma. Recently, experiments have been conducted to elucidate the nature of both the coherent waves and the turbulence. The experiments provide insight into a broad range of interesting linear and nonlinear plasma phenomena and into many of the processes that determine such practical things as plasma heating and confinement. PMID:17753464
Resistive X-point modes in tokamak boundary plasmas
Myra, J. R.; D'Ippolito, D. A.; Xu, X. Q.; Cohen, R. H.
2000-06-01
It is shown that the boundary (edge and scrape-off-layer) plasma in a typical low (L) mode diverted tokamak discharge is unstable to a new class of modes called resistive X-point (RX) modes. The RX mode is a type of resistive ballooning mode that exploits a synergism between resistivity and the magnetic geometry of the X-point region. The RX modes are shown to give robust instabilities at moderate mode numbers, and therefore are expected to be the dominant contributors to turbulent diffusion in the boundary plasma of a diverted tokamak. (c) 2000 American Institute of Physics.
Not Available
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.
Rippling modes in the edge of a Tokamak plasma
NASA Astrophysics Data System (ADS)
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. A computational model for these modes in the cylindrical Tokamak approximation was developed and the linear growth and single helicity quasilinear saturation phases of the rippling modes for parameters appropriate to the edge of a Tokamak plasma was explored. Large parallel heat conduction does not stabilize these mode. Nonlinearly, individual rippling modes are found to saturate by quasilinear 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 B0 convection. It is found that this evolution is terminated by parallel heat conduction.
Experimental observations of driven and intrinsic rotation in tokamak plasmas
NASA Astrophysics Data System (ADS)
Rice, J. E.
2016-08-01
Experimental observations of driven and intrinsic rotation in tokamak plasmas are reviewed. For momentum sources, there is direct drive from neutral beam injection, lower hybrid and ion cyclotron range of frequencies waves (including mode conversion flow drive), as well as indirect \\mathbf{j}× \\mathbf{B} forces from fast ion and electron orbit shifts, and toroidal magnetic field ripple loss. Counteracting rotation drive are sinks, such as from neutral drag and toroidal viscosity. Many of these observations are in agreement with the predictions of neo-classical theory while others are not, and some cases of intrinsic rotation remain puzzling. In contrast to particle and heat fluxes which depend on the relevant diffusivity and convection, there is an additional term in the momentum flux, the residual stress, which can act as the momentum source for intrinsic rotation. This term is independent of the velocity or its gradient, and its divergence constitutes an intrinsic torque. The residual stress, which ultimately responds to the underlying turbulence, depends on the confinement regime and is a complicated function of collisionality, plasma shape, and profiles of density, temperature, pressure and current density. This leads to the rich intrinsic rotation phenomenology. Future areas of study include integration of these many effects, advancement of quantitative explanations for intrinsic rotation and development of strategies for velocity profile control.
Stabilization of tokamak plasma by lithium streams
L.E. Zakharov
2000-08-07
The stabilization theory of free-boundary magnetohydrodynamic instabilities in tokamaks by liquid lithium streams driven by magnetic propulsion is formulated. While the conventional, wall-locked, resistive wall mode can be well suppressed by the flow, a new, stream-locked mode determines the limits of the flow stabilization.
On steady poloidal and toroidal flows in tokamak plasmas
McClements, K. G.
2010-08-15
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{sub {theta}/}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{sub {theta}/}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.
The Science of Spherical Tokamak Plasmas: Progress and Promise
NASA Astrophysics Data System (ADS)
Sykes, Alan
2008-11-01
The talk will summarize the development of the low aspect ratio `Spherical' Tokamak (ST) from early linear magnetic confinement devices, through toroidal pinches, to the emergence of the tokamak in the 1960's. Theoretical predictions given by Peng and Strickler of the exciting physics of extreme low aspect ratio tokamaks (supported by early experiments involving centre rods inserted into existing Rotamaks, Spheromaks and other small-scale experiments), led to the pioneering START experiment at Culham which convincingly demonstrated the potential of the ST concept. There are now many STs world-wide. The largest among these are MA-scale devices NSTX and MAST with plasmas of cross-section comparable to DIII-D and Asdex-Upgrade. The major results include development of start-up methods; the refinement of scaling laws; improved understanding of general tokamak phenomena such as Edge Localised Modes and development of heating and current drive schemes. ST research on over 20 devices has extended the tokamak plasma regime in many ways, notably a factor 4 increase in stable toroidal average beta, and large increases in the Alfven Mach number and ExB flow shear. By exploiting such features, joint experiments with tokamaks of conventional aspect ratio are resolving several key degeneracies of interest to ITER, DEMO and larger future ST devices. Present STs have low toroidal fields sufficient for most physics studies, but for high fusion yield or energy production higher fields are required; importantly, studies on both NSTX and MAST indicate a stronger than expected improvement of performance with toroidal field. Both devices are planning exciting upgrades which feature a considerable increase of toroidal field. Recent designs for a D-T Component Test Facility based on the Spherical Tokamak show the promise of low Tritium consumption and minimum build cost. Such a facility would provide valuable R&D on the scientific and technical issues of fusion power.
Dust-Particle Transport in Tokamak Edge Plasmas
Pigarov, A Y; Krasheninnikov, S I; Soboleva, T K; Rognlien, T D
2005-09-12
Dust particulates in the size range of 10nm-100{micro}m are found in all fusion devices. Such dust can be generated during tokamak operation due to strong plasma/material-surface interactions. Some recent experiments and theoretical estimates indicate that dust particles can provide an important source of impurities in the tokamak plasma. Moreover, dust can be a serious threat to the safety of next-step fusion devices. In this paper, recent experimental observations on dust in fusion devices are reviewed. A physical model for dust transport simulation, and a newly developed code DUSTT, are discussed. The DUSTT code incorporates both dust dynamics due to comprehensive dust-plasma interactions as well as the effects of dust heating, charging, and evaporation. The code tracks test dust particles in realistic plasma backgrounds as provided by edge-plasma transport codes. Results are presented for dust transport in current and next-step tokamaks. The effect of dust on divertor plasma profiles and core plasma contamination is examined.
Advanced tokamak physics experiments on DIII-D
Taylor, T.S.
1998-12-01
Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy confinement time ({tau}{sub E}) and the plasma pressure (or beta {beta}{sub T} = 2 {mu}{sub 0} < p > /B{sub T}{sup 2}) can be achieved in steady-state conditions with high self driven bootstrap current fraction. In addition, effective power exhaust and impurity and particle control is required. Significant progress has been made in experimentally achieving regimes having the required performance in all of these aspects as well as in developing a theoretical understanding of the underlying physics. The authors have extended the duration of high performance ELMing H-mode plasmas with {beta}{sub N} H{sub iop} {approximately} 10 for 5 {tau}{sub E} ({approximately}1 s) and have demonstrated that core transport barriers can be sustained for the entire 5-s neutral beam duration in L-mode plasmas. Recent DIII-D work has advanced the understanding of improved confinement and internal transport barriers in terms of E x B shear stabilization of micro turbulence. With the aim of current profile control in discharges with negative central magnetic shear, they have demonstrated off-axis electron cyclotron current drive for the first time in a tokamak, finding an efficiency above theoretical expectations. MHD stability has been improved through shape optimization, wall stabilization, and modification of the pressure and current density profiles. Heat flux reduction and improved impurity and particle control have been realized through edge/divertor radiation and understanding and utilization of forced scrape off layer flow and divertor baffling.
Problems with the concept of plasma equilibrium in tokamaks
Carreras, B.A.
1992-06-01
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.
Analytic model for coaxial helicity injection in tokamak plasmas
Weening, R. H.
2011-12-15
Using a partial differential equation for the time evolution of the mean-field poloidal magnetic flux that incorporates resistivity {eta} and hyper-resistivity {Lambda} terms, an exact analytic solution is obtained for steady-state coaxial helicity injection (CHI) in force-free large aspect ratio tokamaks. The analytic mean-field Ohm's law model allows for calculation of the tokamak CHI current drive efficiency and the plasma inductances at arbitrary levels of magnetic fluctuations, or dynamo activity. The results of the mean-field model suggest that CHI approaching Ohmic efficiency is only possible in tokamaks when the size of the effective current drive boundary layer, {delta}{identical_to}({Lambda}/{eta}){sup 1/2}, becomes greater than half the size of the plasma, {delta}>a/2, with a the plasma minor radius. The electron thermal diffusivity due to magnetic fluctuation induced transport is obtained from the expression {chi}{sub e}={Lambda}/{mu}{sub 0}d{sub e}{sup 2}, with {mu}{sub 0} the permeability of free space and d{sub e} the electron skin depth, which for typical tokamak fusion plasma parameters is on the order of a millimeter. Thus, the ratio of the energy confinement time to the resistive diffusion time in a tokamak plasma driven by steady-state CHI approaching Ohmic efficiency is shown to be constrained by the relation {tau}{sub E}/{tau}{sub {eta}}<(d{sub e}/a){sup 2}{approx_equal}10{sup -6}. The mean-field model suggests that steady-state CHI can be viewed most simply as a boundary layer of stochastically wandering magnetic field lines.
Plasma filamentation in the Rijnhuizen tokamak RTP
Lopes Cardozo, N.J.; Schueller, F.C.; Barth, C.J.; Chu, C.C.; Pijper, F.J.; Lok, J.; Oomens, A.A.M. )
1994-07-11
Evidence for small scale magnetic structures in the Rijnhuizen tokamak RTP is presented. These are manifest through steps and peaks in the electron temperature and pressure, measured with multiposition Thomson scattering. During central electron cyclotron heating, several filaments of high pressure are found in the power deposition region. They live hundreds of microseconds. Near the sawtooth inversion radius a step'' in the temperature profile occurs. Further out, quasiperiodic structures are observed, in both Ohmic and heated discharges.
A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak.
Ren, J; Zuo, G Z; Hu, J S; Sun, Z; Yang, Q X; Li, J G; Zakharov, L E; Xie, H; Chen, Z X
2015-02-01
A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak-both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST. PMID:25725839
Solenoid-free plasma start-up in spherical tokamaks
NASA Astrophysics Data System (ADS)
Raman, R.; Shevchenko, V. F.
2014-10-01
The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.
Microtearing mode (MTM) turbulence in JIPPT-IIU tokamak plasmas
NASA Astrophysics Data System (ADS)
Hamada, Y.; Watari, T.; Nishizawa, A.; Yamagishi, O.; Narihara, K.; Ida, K.; Kawasumi, Y.; Ido, T.; Kojima, M.; Toi, K.; the JIPPT-IIU Group
2015-04-01
Magnetic, density and potential fluctuations up to 500 kHz at several spatial points have been observed in the core region of JIPPT-IIU tokamak plasmas using a heavy ion beam probe. The frequency spectra of the density and magnetic oscillations are found to be similar, whereas there are large differences in the phase, coherence and frequency dependences deduced from signals at adjacent sample volumes. These differences allow us to ascribe the detected magnetic fluctuations to the microtearing mode (MTM) by simple dispersion relations of the MTM in collisionless and intermediate regimes. The frequency-integrated level of magnetic fluctuations around 150 kHz (100-200 kHz) is \\tilde{{B}}r /Bt ≈ 1× 10-4 , a level high enough for the ergodization of the magnetic surface and enhanced electron heat loss as derived by Rechester and Rosenbluth (1978 Phys. Rev. Lett. 40 38). This level is consistent with the measurements performed using cross-polarization scattering of microwaves in the Tore Supra tokamak. Our results are the first direct experimental verification of the MTM in the core region of tokamak plasmas, which has been recently observed in gyrokinetic simulations using a very fine mesh in tokamak and ST plasmas.
Probing spherical tokamak plasmas using charged fusion products
NASA Astrophysics Data System (ADS)
Boeglin, Werner U.; Perez, Ramona V.; Darrow, Douglass S.; Cecconello, Marco; Klimek, Iwona; Allan, Scott Y.; Akers, Rob J.; Jones, Owen M.; Keeling, David L.; McClements, Ken G.; Scannell, Rory
2015-11-01
The detection of charged fusion products, such as protons and tritons resulting from D(d,p)t reactions, can be used to determine the fusion reaction rate profile in large spherical tokamak plasmas with neutral beam heating. The time resolution of a diagnostic of this type makes it possible to study the slowly-varying beam density profile, as well as rapid changes resulting from MHD instabilities. A 4-channel prototype proton detector (PD) was installed and operated on the MAST spherical tokamak in August/September 2013, and a new 6-channel system for the NSTX-U spherical tokamak is under construction. PD and neutron camera measurements obtained on MAST will be compared with TRANSP calculations, and the design of the new NSTX-U system will be presented, together with the first results from this diagnostic, if available. Supported in part by DOE DE-SC0001157.
Pseudo-MHD ballooning modes in tokamak plasmas
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-MHD{close_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 {alpha} > s{sup 2}(2 {sup 7/3}/9) (r{sub p}/R{sub 0}) or-d{radical}{Beta}/dr > (2{sup 1/6} /3)(s/ R{sub 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.
Tokamak Plasma Flows Induced by Local RF Forces
NASA Astrophysics Data System (ADS)
Chen, Jiale; Gao, Zhe
2015-10-01
The tokamak plasma flows induced by the local radio frequency (RF) forces in the core region are analyzed. The effective components of local RF forces are composed of the momentum absorption term and the resonant parallel momentum transport term (i.e. the parallel component of the resonant ponderomotive forces). Different momentum balance relations are employed to calculate the plasma flows depending on different assumptions of momentum transport. With the RF fields solved from RF simulation codes, the toroidal and poloidal flows by these forces under the lower hybrid current drive and the mode conversion ion cyclotron resonance heating on EAST-like plasmas are evaluated. supported by National Natural Science Foundation of China (Nos. 11405218, 11325524, 11375235 and 11261140327), in part by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB111002, 2013GB112001 and 2013GB112010), and the Program of Fusion Reactor Physics and Digital Tokamak with the CAS “One-Three-Five” Strategic Planning
Two Dimensional Particle Transport in the Cct Tokamak Edge Plasma
NASA Astrophysics Data System (ADS)
Tynan, George Robert
The physics of particle transport in the CCT tokamak plasma edge is studied experimentally in this dissertation. A full poloidal array of Langmuir probes is used to measure the equilibrium plasma and transport properties of the CCT edge plasma during Ohmic and H-mode discharges. During Ohmic L-mode, the equilibrium plasma density and electron temperature are found to vary on a magnetic flux surface. The equilibrium plasma distribution coincides with the distribution of particle transport. Inside the last closed flux surface, convective processes dominate particle transport. Several large convective cells are observed near the limiter radius. At and beyond the limiter radius, turbulent transport is significant. The turbulence appears to be driven by the convective plasma flows. In Ohmic L-mode-like discharges, plasma transport occurs predominantly through the low-field region of the tokamak with local bad curvature. The convective cells are destroyed at the L-H transition and replaced with a more poloidally symmetric, radially narrow jet of plasma flow at the limiter radius. The jet effectively isolates the plasma core from the scrape -off layer. The turbulence associated with the convective cells is reduced across the edge region. Radial particle transport across the limiter radius is thus inhibited and the global particle confinement is increased. The available data suggest that the residual H-mode particle transport is more poloidally symmetric.
Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST
Xu, X Q
2007-11-09
We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D ({psi}, {theta}, {epsilon}, {mu}) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices.
Observation of Energetic Particle Driven Modes Relevant to Advanced Tokamak Regimes
R. Nazikian; B. Alper; H.L. Berk; D. Borba; C. Boswell; R.V. Budny; K.H. Burrell; C.Z. Cheng; E.J. Doyle; E. Edlund; R.J. Fonck; A. Fukuyama; N.N. Gorelenkov; C.M. Greenfield; D.J. Gupta; M. Ishikawa; R.J. Jayakumar; G.J. Kramer; Y. Kusama; R.J. La Haye; G.R. McKee; W.A. Peebles; S.D. Pinches; M. Porkolab; J. Rapp; T.L. Rhodes; S.E. Sharapov; K. Shinohara; J.A. Snipes; W.M. Solomon; E.J. Strait; M. Takechi; M.A. Van Zeeland; W.P. West; K.L. Wong; S. Wukitch; L. Zeng
2004-10-21
Measurements of high-frequency oscillations in JET [Joint European Torus], JT-60U, Alcator C-Mod, DIII-D, and TFTR [Tokamak Fusion Test Reactor] plasmas are contributing to a new understanding of fast ion-driven instabilities relevant to Advanced Tokamak (AT) regimes. A model based on the transition from a cylindrical-like frequency-chirping mode to the Toroidal Alfven Eigenmode (TAE) has successfully encompassed many of the characteristics seen in experiments. In a surprising development, the use of internal density fluctuation diagnostics has revealed many more modes than has been detected on edge magnetic probes. A corollary discovery is the observation of modes excited by fast particles traveling well below the Alfven velocity. These observations open up new opportunities for investigating a ''sea of Alfven Eigenmodes'' in present-scale experiments, and highlight the need for core fluctuation and fast ion measurements in a future burning-plasma experiment.
Public Data Set: Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup
Hinson, Edward T. [University of Wisconsin-Madison] (ORCID:000000019713140X); Barr, Jayson L. [University of Wisconsin-Madison] (ORCID:0000000177685931); Bongard, Michael W. [University of Wisconsin-Madison] (ORCID:0000000231609746); Burke, Marcus G. [University of Wisconsin-Madison] (ORCID:0000000176193724); Fonck, Raymond J. [University of Wisconsin-Madison] (ORCID:0000000294386762); Perry, Justin M. [University of Wisconsin-Madison] (ORCID:0000000171228609)
2016-05-31
This data set contains openly-documented, machine readable digital research data corresponding to figures published in E.T. Hinson et al., 'Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup,' Physics of Plasmas 23, 052515 (2016).
Toroidally symmetric plasma vortex at tokamak divertor null point
NASA Astrophysics Data System (ADS)
Umansky, M. V.; Ryutov, D. D.
2016-03-01
Reduced MHD equations are used for studying toroidally symmetric plasma dynamics near the divertor null point. Numerical solution of these equations exhibits a plasma vortex localized at the null point with the time-evolution defined by interplay of the curvature drive, magnetic restoring force, and dissipation. Convective motion is easier to achieve for a second-order null (snowflake) divertor than for a regular x-point configuration, and the size of the convection zone in a snowflake configuration grows with plasma pressure at the null point. The trends in simulations are consistent with tokamak experiments which indicate the presence of enhanced transport at the null point.
Formation and Stability of Impurity "snakes" in Tokamak Plasmas
L. Delgado-Aparicio, et. al.
2013-01-28
New observations of the formation and dynamics of long-lived impurity-induced helical "snake" modes in tokamak plasmas have recently been carried-out on Alcator C-Mod. The snakes form as an asymmetry in the impurity ion density that undergoes a seamless transition from a small helically displaced density to a large crescent-shaped helical structure inside q < 1, with a regularly sawtoothing core. The observations show that the conditions for the formation and persistence of a snake cannot be explained by plasma pressure alone. Instead, many features arise naturally from nonlinear interactions in a 3D MHD model that separately evolves the plasma density and temperature
Plasma flows in scrape-off layer of Aditya tokamak
Sangwan, Deepak; Jha, Ratneshwar; Brotankova, Jana; Gopalkrishna, M. V.
2012-09-15
The magnetized Mach probe is used to make measurement of plasma flows in the scrape-off layer of the Aditya tokamak [R. Jha et al., Plasma Phys. Controlled Fusion 51, 095010 (2009)]. This probe is further used to measure dependencies of Mach number on local plasma densities and radial distances of the probe in the scrape-off layer. The measured Mach number has contributions from E Multiplication-Sign B drift, Pfrisch-Schlueter, and transport driven flows. We have determined that the toroidal flow is towards the ion side of the limiter and the poloidal flow direction is towards the contact of the last closed flux surface with the limiter.
Influence of plasma surface interactions on tokamak startup
Goswami, Rajiv
2013-08-15
The startup phase of a tokamak is a complex phenomenon involving burnthrough of the low-Z impurities and rampup of I{sub p}, the plasma current. The design considerations of a tokamak are closely connected with the startup modeling. Plasma evolution is analysed using a zero-dimensional model. The particle and energy balance is considered of two subclasses of plasmas which are penetrable by neutral gas, together with another component, neutrals trapped in the wall. The first subclass includes plasmas being penetrated by slow neutrals of (∼few eV) temperature. The second includes plasmas being penetrated only by fast neutrals having a temperature comparable to that of the ions. The impact of impurities on energy balance is considered through their generation by ion induced desorption of adsorbed oxygen on the first wall and physical and chemical sputtering of carbon. The paper demonstrates self-consistently that the evolution of initial phase of the discharge is intimately linked to the condition of the plasma facing components (PFCs) and the resultant plasma surface interactions.
Drift-wave fluctuation in an inviscid tokamak plasma
NASA Astrophysics Data System (ADS)
Yang, Jian-Rong; Mao, Jie-Jian; Tang, Xiao-Yan
2013-11-01
In order to describe the characterization of resistive drift-wave fluctuation in a tokamak plasma, a coupled inviscid two-dimensional Hasegawa—Wakatani model is investigated. Two groups of new analytic solutions with and without phase shift between the fluctuant density and the fluctuant potential are obtained by using the special function transformation method. It is demonstrated that the fluctuant potential shares similar spatio—temporal variations with the density. It is found from the solutions without phase shift that the effect of the diffusion and adiabaticity on the fluctuant density is quite complex, and that the fluctuation may be controlled through the adiabaticity and diffusion. By using the typical parameters in the quasi-adiabatic regime in the solutions with phase shift, it is shown that the density gradient becomes larger as the contours become dense toward the plasma edge and the contours have irregular structures, which reveal the nonuniform distribution in the tokamak edge.
'Snowflake' H Mode in a Tokamak Plasma
Piras, F.; Coda, S.; Duval, B. P.; Labit, B.; Marki, J.; Moret, J.-M.; Pitzschke, A.; Sauter, O.; Medvedev, S. Yu.
2010-10-08
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 ({Delta}W{sub ELM}/W{sub p}) compared to an identically shaped, conventional single-null diverted H mode. Enhanced stability of mid- to high-toroidal-mode-number ideal modes is consistent with the different snowflake ELM phenomenology. The capability of the snowflake to redistribute the edge power on the additional strike points has been confirmed experimentally.
Stochastic Acceleration of Dust Particles in Tokamak Edge Plasmas
Marmolino, C.; De Angelis, U.; Ivlev, A. V.; Morfill, G. E.
2008-10-15
Stochastic heating of dust particles resulting from dust charge fluctuations is considered in the conditions of the scrape-off-layer (SOL) in tokamak plasmas. It is shown that kinetic energies corresponding to velocities of {approx_equal}Km/s can be reached in times of order {approx_equal}1 ms by micron-size dust particles interacting with a background of stochastically heated nano-size dust particles.
Momentum Injection in Tokamak Plasmas and Transitions to Reduced Transport
Parra, F. I.; Highcock, E. G.; Schekochihin, A. A.; Barnes, M.
2011-03-18
The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.
Edge Plasma Studies and Related Diagnostics on CASTOR Tokamak
Hron, M.; Stockel, J.; Duran, I.; Panek, R.; Adamek, J.; Weinzettl, V.
2006-12-04
In this contribution, two sets of measurements using a full poloidal array of Langmuir probes in the scrape-off layer of the CASTOR tokamak are described. First, results obtained with edge plasma biasing show creation of convective cells that cause radial transport due to ExB drift. Next, the analysis of the turbulence behaviour in standard ohmic discharges shows the presence of a spatially periodical mode with mode number equal to the edge safety factor q.
Tokamak plasma current disruption infrared control system
Kugel, Henry W.; Ulrickson, Michael
1987-01-01
In a magnetic plasma confinment device having an inner toroidal limiter mounted on an inner wall of a plasma containment vessel, an arrangement is provided for monitoring vertical temperature profiles of the limiter. The temperature profiles are taken at brief time intervals, in a time scan fashion. The time scans of the vertical temperature profile are continuously monitored to detect the presence of a peaked temperature excursion, which, according to the present invention, is a precursor of a subsequent major plasma disruption. A fast scan of the temperature profile is made so as to provide a time interval in real time prior to the major plasma disruption, such that corrective action can be taken to reduce the harmful effects of the plasma disruption.
Development on JET of advanced tokamak operations for ITER
NASA Astrophysics Data System (ADS)
Tuccillo, A. A.; Crisanti, F.; Litaudon, X.; Baranov, Yu. F.; Becoulet, A.; Becoulet, M.; Bertalot, L.; Castaldo, C.; Challis, C. D.; Cesario, R.; DeBaar, M. R.; de Vries, P. C.; Esposito, B.; Frigione, D.; Garzotti, L.; Giovannozzi, E.; Giroud, C.; Gorini, G.; Gormezano, C.; Hawkes, N. C.; Hobirk, J.; Imbeaux, F.; Joffrin, E.; Lomas, P. J.; Mailloux, J.; Mantica, P.; Mantsinen, M. J.; Mazon, D.; Moreau, D.; Murari, A.; Pericoli-Ridolfini, V.; Rimini, F.; Sips, A. C. C.; Sozzi, C.; Tudisco, O.; Van Eester, D.; Zastrow, K.-D.; work-programme contributors, JET-EFDA
2006-02-01
Recent research on advanced tokamak in JET has focused on scenarios with both monotonic and reversed shear q-profiles having plasma parameters as relevant as possible for extrapolation to ITER. Wide internal transport barriers (ITBs), r/a ~ 0.7, are formed at ITER relevant triangularity δ ~ 0.45 and moderate plasma current, IP = 1.5-2.5 MA, with ne/nG ~ 60% when ELMs are moderated by Ne injection. At higher current (IP <= 3.5 MA, δ ~ 0.25) wide ITBs sitting at r/a >= 0.5, in the positive shear region, have been developed. Generally MHD events terminate these barriers otherwise limited in strength by power availability. ITBs with core density close to Greenwald value, Te ~ Ti and low toroidal rotation (4 times lower than standard ITBs) are obtained in plasma target preformed by opportune timing of lower hybrid current drive (LHCD), pellet injection and a small amount of NBI power. Wide ITBs, r/a ~ 0.6, of moderate strength, can be sustained without impurities accumulation for a time close to neoclassical resistive time in 3 T/1.8 MA discharges that exhibit reversed magnetic shear profiles and type-III ELMy edge. These discharges have been extended to the maximum duration allowed by JET subsystems (20 s) bringing to the record of injected energy in a JET discharge: E ~ 330 MJ. Portability of ITB physics has been addressed through dedicated similarity experiments. The ITB is identified as a layer of reduced diffusivity studying the propagation of the heat wave generated by modulating the ICRF mode conversion (MC) electron heating. Impressive results, QDT ~ 0.25, are obtained in these deuterium discharges with 3He minority when the MC layer is located in the core. The ion behaviour has been investigated in pure LHCD electron ITBs optimizing the 3He minority concentration for direct ion heating. Preliminary results of particle transport, studied via injection of a trace of tritium and an Ar-Ne mixture, will be presented.
Elements of Neoclassical Theory and Plasma Rotation in a Tokamak
NASA Astrophysics Data System (ADS)
Smolyakov, A.
2015-12-01
The following sections are included: * Introduction * Quasineutrality condition * Diffusion in fully ionized magnetized plasma and automatic ambipolarity * Toroidal geometry and neoclassical diffusion * Diffusion and ambipolarity in toroidal plasmas * Ambipolarity and equilibrium poloidal rotation * Ambipolarity paradox and damping of poloidal rotation * Neoclassical plasma inertia * Oscillatory modes of poloidal plasma rotation * Dynamics of the toroidal momentum * Momentum diffusion in strongly collisional, short mean free path regime * Diffusion of toroidal momentum in the weak collision (banana) regime * Toroidal momentum diffusion and momentum damping from drift-kinetic theory and fluid moment equations * Comments on non-axisymmetric effects * Summary * Acknowledgments * Appendix: Trapped (banana) particles and collisionality regimes in a tokamak * Appendix: Hierarchy of moment equations * Appendix: Plasma viscosity tensor in the magnetic field: parallel viscosity, gyroviscosity, and perpendicular viscosity * Appendix: Closure relations for the flux surface averaged parallel viscosity in neoclassical (banana and plateau) regimes * References
Offline Development of Plasma Boundary Controllers for the KSTAR Tokamak
NASA Astrophysics Data System (ADS)
Ballinger, S.; Eidietis, N. W.; Humphreys, D. A.; Hyatt, A. W.; Welander, A. S.; Hahn, S. H.
2014-10-01
The KSTAR TokSys tokamak simulator, implemented in Matlab®/Simulink, has been extended to include a plasma boundary control system to allow automated offline tuning of shape control feedback loops. Offline control development minimizes resources expended tuning controllers during actual run time, and automated tuning is desirable in order to optimize the large number of shape control gains. The new simulation includes simplified versions of the rtEFIT/Isoflux controller used in the KSTAR plasma control system, allowing full-closed-loop analysis of the plasma shape control. Results presented include application of robust design methods to optimizing control of KSTAR's plasma boundary, and analysis to understand observed differences in boundary control between KSTAR and other superconducting devices. Work supported in part by the National Undergraduate Fellowship Program in Plasma Physics and Fusion Energy Sciences and the US Department of Energy under DE-FC02-04ER54698.
Liquid gallium jet-plasma interaction studies in ISTTOK tokamak
NASA Astrophysics Data System (ADS)
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.
2009-06-01
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.
Eikonal waves, caustics and mode conversion in tokamak plasmas
NASA Astrophysics Data System (ADS)
Jaun, A.; Tracy, E. R.; Kaufman, A. N.
2007-01-01
Ray optics is used to model the propagation of short electromagnetic plasma waves in toroidal geometry. The new RAYCON code evolves each ray independently in phase space, together with its amplitude, phase and focusing tensor to describe the transport of power along the ray. Particular emphasis is laid on caustics and mode conversion layers, where a linear phenomenon splits a single incoming ray into two. The complete mode conversion algorithm is described and tested for the first time, using the two space dimensions that are relevant in a tokamak. Applications are shown, using a cold plasma model to account for mode conversion at the ion-hybrid resonance in the Joint European Torus.
Continuum kinetic modeling of the tokamak plasma edge
NASA Astrophysics Data System (ADS)
Dorf, M. A.; Dorr, M. R.; Hittinger, J. A.; Cohen, R. H.; Rognlien, T. D.
2016-05-01
The first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalous radial transport.
Oxygen impurity radiation from Tokamak-like plasmas
NASA Technical Reports Server (NTRS)
Rogerson, J. E.; Davis, J.; Jacobs, V. L.
1977-01-01
We have constructed a nonhydrodynamic coronal model for calculating radiation from impurity atoms in a heated plasma. Some recent developments in the calculation of dielectronic recombination rate coefficients and collisional excitation rate coefficients are included. The model is applied to oxygen impurity radiation during the first few milliseconds of a TFR Tokamak plasma discharge, and good agreement with experimental results is obtained. Estimates of total line and continuum radiation from the oxygen impurity are given. It is shown that impurity radiation represents a considerable energy loss.
Turbulence studies in Tokamak boundary plasmas with realistic divertor geometry
Xu, X.Q.
1998-10-14
Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT [1] and the linearized shooting code BAL[2] to study turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant, resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters.
Diamagnetic thresholds for sawtooth cycling in tokamak plasmas
NASA Astrophysics Data System (ADS)
Halpern, Federico D.; Lütjens, Hinrich; Luciani, Jean-François
2011-10-01
The cycling dynamics of the internal kink mode, which drives sawtooth oscillations in tokamak plasmas, is studied using the three dimensional, non-linear magnetohydrodynamic (MHD) code XTOR-2F [H. Lütjens and J.-F. Luciani, J. Comput. Phys. 229, 8130 (2010)]. It is found that sawtooth cycling, which is characterized by quiescent ramps and fast crashes in the experiment, can be recovered in two-fluid MHD provided that a criterion of diamagnetic stabilization is fulfilled. The simulation results indicate that diamagnetic effects alone may be sufficient to drive sawteeth with complete magnetic reconnection in high temperature Ohmic plasmas.
OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM
BURRELL,KH
2002-11-01
OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, the authors have made significant progress in developing the building blocks needed for AT operation: (1) the authors have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {le} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. They have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiation power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet
Advanced Tokamak Regimes in Alcator C-Mod with Lower Hybrid Current Drive
NASA Astrophysics Data System (ADS)
Parker, R.; Bonoli, P.; Gwinn, D.; Hutchinson, I.; Porkolab, M.; Ramos, J.; Bernabei, S.; Hosea, J.; Wilson, R.
1999-11-01
Alcator C-Mod has been proposed as a test-bed for developing advanced tokamak scenarios owing to its strong shaping, relatively long pulse length capability at moderate field, e.g. t ~ L/R at B = 5T and T_eo ~ 7keV, and the availability of strong ICRF heating. We plan to exploit this capability by installing up to 4 MW RF power at 4.6 GHz for efficient off-axis current drive by lower hybrid waves. By launching LH waves with a grill whose n_xx spectrum can be dynamically controlled over the range 2 < n_xx < 3.5, the driven current profile can be modified so that, when combined with bootstrap current in high ɛβ_pol regimes, q_min > 2. Such reversed or nearly zero shear regimes have already been proposed as the basis of an advanced tokamak burning-plasma experiment-ATBX (M. Porkolab et al, IAEA-CN-69/FTP/13, IAEA,Yokohama 1998.), and could provide the basis for a demonstration power reactor. Theoretical and experimental basis for this advanced tokamak research program on C-Mod, including design of the lower hybrid coupler, its spectrum and current drive capabilities will be presented.
Comments on experimental results of energy confinement of tokamak plasmas
Chu, T.K.
1989-04-01
The results of energy-confinement experiments on steady-state tokamak plasmas are examined. For plasmas with auxiliary heating, an analysis based on the heat diffusion equation is used to define heat confinement time (the incremental energy confinement time). For ohmically sustained plasmas, experiments show that the onset of the saturation regime of energy confinement, marfeing, detachment, and disruption are marked by distinct values of the parameter /bar n//sub e///bar j/. The confinement results of the two types of experiments can be described by a single surface in 3-dimensional space spanned by the plasma energy, the heating power, and the plasma density: the incremental energy confinement time /tau//sub inc/ = ..delta..W/..delta..P is the correct concept for describing results of heat confinement in a heating experiment; the commonly used energy confinement time defined by /tau//sub E/ = W/P is not. A further examination shows that the change of edge parameters, as characterized by the change of the effective collision frequency ..nu../sub e/*, governs the change of confinement properties. The totality of the results of tokamak experiments on energy confinement appears to support a hypothesis that energy transport is determined by the preservation of the pressure gradient scale length. 70 refs., 6 figs., 1 tab.
NASA Astrophysics Data System (ADS)
Zakharov, Leonid E.; Li, Xujing
2015-06-01
This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97-104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.
Zakharov, Leonid E.; Li, Xujing
2015-06-15
This paper formulates the Tokamak Magneto-Hydrodynamics (TMHD), initially outlined by X. Li and L. E. Zakharov [Plasma Science and Technology 17(2), 97–104 (2015)] for proper simulations of macroscopic plasma dynamics. The simplest set of magneto-hydrodynamics equations, sufficient for disruption modeling and extendable to more refined physics, is explained in detail. First, the TMHD introduces to 3-D simulations the Reference Magnetic Coordinates (RMC), which are aligned with the magnetic field in the best possible way. The numerical implementation of RMC is adaptive grids. Being consistent with the high anisotropy of the tokamak plasma, RMC allow simulations at realistic, very high plasma electric conductivity. Second, the TMHD splits the equation of motion into an equilibrium equation and the plasma advancing equation. This resolves the 4 decade old problem of Courant limitations of the time step in existing, plasma inertia driven numerical codes. The splitting allows disruption simulations on a relatively slow time scale in comparison with the fast time of ideal MHD instabilities. A new, efficient numerical scheme is proposed for TMHD.
Molecular emission in the edge plasma of T-10 tokamak
Zimin, A. M.; Krupin, V. A.; Troynov, V. I.; Klyuchnikov, L. A.
2015-12-15
The experiments on recording molecular emission in the edge plasma of the T-10 tokamak are described. To obtain reliable spectra with sufficient spectral, temporal, and spatial resolution, the optical circuit is optimized for various experimental conditions. Typical spectra measured in two sections of the tokamak are shown. It is shown that, upon varying the parameters of the discharge, the molecular spectrum not only changes significantly in intensity but also undergoes a qualitative change in the rotational and vibrational structure. For a detailed analysis, we use the Fulcher-α system (d{sup 3}Π{sub u}–a{sup 3}Σ{sub g}{sup +}) of deuterium in the wavelength range from 590 to 640 nm. The rotational temperatures of ground state X{sup 1}Σ{sub g}{sup +} and upper excited state d{sup 3}Π{sub u} are estimated by the measured spectra.
Neoclassical momentum transport in an impure rotating tokamak plasma
Newton, S.; Helander, P.
2006-01-15
It is widely believed that transport barriers in tokamak plasmas are caused by radial electric-field shear, which is governed by angular momentum transport. Turbulence is suppressed in the barrier, and ion thermal transport is comparable to the neoclassical prediction, but experimentally angular momentum transport has remained anomalous. With this motivation, the collisional transport matrix is calculated for a low collisionality plasma with collisional impurity ions. The bulk plasma toroidal rotation velocity is taken to be subsonic, but heavy impurities undergo poloidal redistribution due to the centrifugal force. The impurities give rise to off-diagonal terms in the transport matrix, which cause the plasma to rotate spontaneously. At conventional aspect ratio, poloidal impurity redistribution increases the angular momentum flux by a factor up to {epsilon}{sup -3/2} over previous predictions, making it comparable to the 'banana' regime heat flux. The flux is primarily driven by radial pressure and temperature gradients.
Tokamak plasma current disruption infrared control system
Kugel, H.W.; Ulrickson, M.
1984-04-16
This invention is directed to the diagnosis and detection of gross or macroinstabilities in a magnetically-confined fusion plasma device. Detection is performed in real time, and is prompt such that correction of the instability can be initiated in a timely fashion.
A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak
Ren, J.; Zuo, G. Z.; Hu, J. S.; Sun, Z.; Yang, Q. X.; Li, J. G.; Xie, H.; Chen, Z. X.; Zakharov, L. E.
2015-02-15
A program involving the extensive and systematic use of lithium (Li) as a “first,” or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.
A flowing liquid lithium limiter for the Experimental Advanced Superconducting Tokamak
NASA Astrophysics Data System (ADS)
Ren, J.; Zuo, G. Z.; Hu, J. S.; Sun, Z.; Yang, Q. X.; Li, J. G.; Zakharov, L. E.; Xie, H.; Chen, Z. X.
2015-02-01
A program involving the extensive and systematic use of lithium (Li) as a "first," or plasma-facing, surface in Tokamak fusion research devices located at Institute of Plasma Physics, Chinese Academy of Sciences, was started in 2009. Many remarkable results have been obtained by the application of Li coatings in Experimental Advanced Superconducting Tokamak (EAST) and liquid Li limiters in the HT-7 Tokamak—both located at the institute. In furtherance of the lithium program, a flowing liquid lithium (FLiLi) limiter system has been designed and manufactured for EAST. The design of the FLiLi limiter is based on the concept of a thin flowing film which was previously tested in HT-7. Exploiting the capabilities of the existing material and plasma evaluation system on EAST, the limiter will be pre-wetted with Li and mechanically translated to the edge of EAST during plasma discharges. The limiter will employ a novel electro-magnetic pump which is designed to drive liquid Li flow from a collector at the bottom of limiter into a distributor at its top, and thus supply a continuously flowing liquid Li film to the wetted plasma-facing surface. This paper focuses on the major design elements of the FLiLi limiter. In addition, a simulation of incoming heat flux has shown that the distribution of heat flux on the limiter surface is acceptable for a future test of power extraction on EAST.
Magnetized plasma flow injection into tokamak and high-beta compact torus plasmas
NASA Astrophysics Data System (ADS)
Matsunaga, Hiroyuki; Komoriya, Yuuki; Tazawa, Hiroyasu; Asai, Tomohiko; Takahashi, Tsutomu; Steinhauer, Loren; Itagaki, Hirotomo; Onchi, Takumi; Hirose, Akira
2010-11-01
As an application of a magnetized coaxial plasma gun (MCPG), magnetic helicity injection via injection of a highly elongated compact torus (magnetized plasma flow: MPF) has been conducted on both tokamak and field-reversed configuration (FRC) plasmas. The injected plasmoid has significant amounts of helicity and particle contents and has been proposed as a fueling and a current drive method for various torus systems. In the FRC, MPF is expected to generate partially spherical tokamak like FRC equilibrium by injecting a significant amount of magnetic helicity. As a circumstantial evidence of the modified equilibrium, suppressed rotational instability with toroidal mode number n = 2. MPF injection experiments have also been applied to the STOR-M tokamak as a start-up and current drive method. Differences in the responses of targets especially relation with beta value and the self-organization feature will be studied.
Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations
Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.
1986-06-01
Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.
Modeling plasma/material interactions during a tokamak disruption
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.
Forbidden line emission from highly ionized atoms in tokamak plasmas
NASA Technical Reports Server (NTRS)
Feldman, U.; Doschek, G. A.; Bhatia, A. K.
1982-01-01
Considerable interest in the observation of forbidden spectral lines from highly ionized atoms in tokamak plasmas is related to the significance of such observations for plasma diagnostic applications. Atomic data for the elements Ti Cr, Mn, Fe, Ni, and Kr have been published by Feldman et al. (1980) and Bhatia et al. (1980). The present investigation is concerned with collisional excitation rate coefficients and radiative decay rates, which are interpolated for ions of elements between calcium, and krypton and for levels of the 2s2 2pk, 2s 2p(k+1), and 2p(k+2) configurations, and for the O I, N I, C I, B I, and Be I isoelectronic sequences. The provided interpolated atomic data can be employed to calculate level populations and relative line intensities for ions of the considered sequences, taking into account levels of the stated configurations. Important plasma diagnostic information provided by the forbidden lines includes the ion temperature
On plasma rotation induced by waves in tokamaks
Guan, Xiaoyin; Dodin, I. Y.; Fisch, N. J.; Qin, Hong; Department of Modern Physics, University of Science and Technology of China, Hefei, Anhui 230026 ; Liu, Jian
2013-10-15
The momentum conservation for resonant wave-particle interactions, now proven rigorously and for general settings, is applied to explain in simple terms how tokamak plasma is spun up by the wave momentum perpendicular to the dc magnetic field. The perpendicular momentum is passed through resonant particles to the dc field and, giving rise to the radial electric field, is accumulated as a Poynting flux; the bulk plasma is then accelerated up to the electric drift velocity proportional to that flux, independently of collisions. The presence of this collisionless acceleration mechanism permits varying the ratio of the average kinetic momentum absorbed by the resonant-particle and bulk distributions depending on the orientation of the wave vector. Both toroidal and poloidal forces are calculated, and a fluid model is presented that yields the plasma velocity at equilibrium.
Halo current diagnostic system of experimental advanced superconducting tokamak.
Chen, D L; Shen, B; Granetz, R S; Sun, Y; Qian, J P; Wang, Y; Xiao, B J
2015-10-01
The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well. PMID:26520954
Halo current diagnostic system of experimental advanced superconducting tokamak
Chen, D. L.; Shen, B.; Sun, Y.; Qian, J. P. Wang, Y.; Xiao, B. J.; Granetz, R. S.
2015-10-15
The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.
Electromagnetic effects on trace impurity transport in tokamak plasmas
Hein, T.; Angioni, C.
2010-01-15
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 ExB transport. A significant impact on the impurity transport due to an increase in the plasma normalized pressure parameter beta is observed in the case of ITG modes, while for TEM the overall effect remains weak. In realistic conditions of high beta plasmas in the high confinement (H-) mode with dominant ITG turbulence, the impurity diffusivity is found to decrease with increasing beta 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 beta, particularly at low impurity charge, due to a compensation between the different off-diagonal contributions.
Kinetic theory of plasma adiabatic major radius compression in tokamaks
NASA Astrophysics Data System (ADS)
Gorelenkova, M. V.; Gorelenkov, N. N.; Azizov, E. A.; Romannikov, A. N.; Herrmann, H. W.
1998-05-01
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'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.
Information content of transient synchrotron radiation in tokamak plasmas
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.
Gyrokinetic simulation of isotope scaling in tokamak plasmas
Lee, W.W.; Santoro, R.A.
1995-07-01
A three-dimensional global gyrokinetic particle code in toroidal geometry has been used for investigating the transport properties of ion temperature gradient (ITG) drift instabilities in tokamak plasmas. Using the isotopes of hydrogen (H{sup +}), deuterium (D{sup +}) and tritium (T{sup +}), we have found that, under otherwise identical conditions, there exists a favorable isotope scaling for the ion thermal diffusivity, i.e., Xi decreases with mass. Such a scaling, which exists both at the saturation of the instability and also at the nonlinear steady state, can be understood from the resulting wavenumber and frequency spectra.
The study of heat flux for disruption on experimental advanced superconducting tokamak
NASA Astrophysics Data System (ADS)
Yang, Zhendong; Fang, Jianan; Gong, Xianzu; Gan, Kaifu; Luo, Jiarong; Zhao, Hailin; Cui, Zhixue; Zhang, Bin; Chen, Meiwen
2016-05-01
Disruption of the plasma is one of the most dangerous instabilities in tokamak. During the disruption, most of the plasma thermal energy is lost, which causes damages to the plasma facing components. Infrared (IR) camera is an effective tool to detect the temperature distribution on the first wall, and the energy deposited on the first wall can be calculated from the surface temperature profile measured by the IR camera. This paper concentrates on the characteristics of heat flux distribution onto the first wall under different disruptions, including the minor disruption and the vertical displacement events (VDE) disruption. Several minor disruptions have been observed before the major disruption under the high plasma density in experimental advanced superconducting tokamak. During the minor disruption, the heat fluxes are mainly deposited on the upper/lower divertors. The magnetic configuration prior to the minor disruption is a lower single null with the radial distance between the two separatrices in the outer midplane dRsep = -2 cm, while it changes to upper single null (dRsep = 1.4 cm) during the minor disruption. As for the VDE disruption, the spatial distribution of heat flux exhibits strong toroidal and radial nonuniformity, and the maximum heat flux received on the dome plate can be up to 11 MW/m2.
Structure of micro-instabilities in tokamak plasmas: Stiff transport or plasma eruptions?
Dickinson, D.
2014-01-15
Solutions to a model 2D eigenmode equation describing micro-instabilities in tokamak plasmas are presented that demonstrate a sensitivity of the mode structure and stability to plasma profiles. In narrow regions of parameter space, with special plasma profiles, a maximally unstable mode is found that balloons on the outboard side of the tokamak. This corresponds to the conventional picture of a ballooning mode. However, for most profiles, this mode cannot exist, and instead, a more stable mode is found that balloons closer to the top or bottom of the plasma. Good quantitative agreement with a 1D ballooning analysis is found, provided the constraints associated with higher order profile effects, often neglected, are taken into account. A sudden transition from this general mode to the more unstable ballooning mode can occur for a critical flow shear, providing a candidate model for why some experiments observe small plasma eruptions (Edge Localised Modes, or ELMs) in place of large Type I ELMs.
THz time-domain spectroscopy for tokamak plasma diagnostics
NASA Astrophysics Data System (ADS)
Causa, F.; Zerbini, M.; Johnston, M.; Buratti, P.; Doria, A.; Gabellieri, L.; Gallerano, G. P.; Giovenale, E.; Pacella, D.; Romano, A.; Tuccillo, A. A.; Tudisco, O.
2014-08-01
The technology is now becoming mature for diagnostics using large portions of the electromagnetic spectrum simultaneously, in the form of THz pulses. THz radiation-based techniques have become feasible for a variety of applications, e.g., spectroscopy, imaging for security, medicine and pharmaceutical industry. In particular, time-domain spectroscopy (TDS) is now being used also for plasma diagnostics in various fields of application. This technique is promising also for plasmas for fusion applications, where plasma characteristics are non-uniform and/or evolve during the discharge This is because THz pulses produced with femtosecond mode-locked lasers conveniently span the spectrum above and below the plasma frequency and, thus, can be used as very sensitive and versatile probes of widely varying plasma parameters. The short pulse duration permits time resolving plasma characteristics while the large frequency span permits a large dynamic range. The focus of this work is to present preliminary experimental and simulation results demonstrating that THz TDS can be realistically adapted as a versatile tokamak plasma diagnostic technique.
THz time-domain spectroscopy for tokamak plasma diagnostics
Causa, F.; Zerbini, M.; Buratti, P.; Gabellieri, L.; Pacella, D.; Romano, A.; Tuccillo, A. A.; Tudisco, O.; Johnston, M.; Doria, A.; Gallerano, G. P.; Giovenale, E.
2014-08-21
The technology is now becoming mature for diagnostics using large portions of the electromagnetic spectrum simultaneously, in the form of THz pulses. THz radiation-based techniques have become feasible for a variety of applications, e.g., spectroscopy, imaging for security, medicine and pharmaceutical industry. In particular, time-domain spectroscopy (TDS) is now being used also for plasma diagnostics in various fields of application. This technique is promising also for plasmas for fusion applications, where plasma characteristics are non-uniform and/or evolve during the discharge This is because THz pulses produced with femtosecond mode-locked lasers conveniently span the spectrum above and below the plasma frequency and, thus, can be used as very sensitive and versatile probes of widely varying plasma parameters. The short pulse duration permits time resolving plasma characteristics while the large frequency span permits a large dynamic range. The focus of this work is to present preliminary experimental and simulation results demonstrating that THz TDS can be realistically adapted as a versatile tokamak plasma diagnostic technique.
ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies
NASA Astrophysics Data System (ADS)
Whyte, Dennis; ADX Team
2015-11-01
The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.
Continuum Kinetic Modeling of the Tokamak Plasma Edge
NASA Astrophysics Data System (ADS)
Dorf, Mikhail
2015-11-01
The problem of edge plasma transport provides substantial challenges for analytical or numerical analysis due to (a) complex magnetic geometry including both open and closed magnetic field lines B, (b) steep radial gradients comparable to ion drift-orbit excursions, and (c) a variation in the collision mean-free path along B from long to short compared to the magnetic connection length. Here, the first 4D continuum drift-kinetic transport simulations that span the magnetic separatrix of a tokamak are presented, motivated in part by the success of continuum kinetic codes for core physics and in part by the potential for high accuracy. The calculations include fully-nonlinear Fokker-Plank collisions and electrostatic potential variations. The problem of intrinsic toroidal rotation driven by ion orbit loss is addressed in detail. The code, COGENT, developed by the Edge Simulation Laboratory collaboration, is distinguished by a fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasma transport and the complex magnetic X-point divertor geometry with high accuracy. Previously, successful performance of high-order algorithms has been demonstrated in a simpler closed magnetic-flux-surface geometry for the problems of neoclassical transport and collisionless relaxation of geodesic acoustic modes in a tokamak pedestal, including the effects of a strong radial electric field under H-mode conditions. Work performed for USDOE, at LLNL under contract DE-AC52-07NA27344.
Transport timescale calculations of sawteeth and helical structures in non-circular tokamak plasmas
NASA Astrophysics Data System (ADS)
Jardin, Stephen; Ferraro, Nate; Breslau, Josh; Chen, Jin
2012-10-01
We present results of using the implicit 3D MHD code M3D-C^1 [1,2] to perform 3D nonlinear magnetohydrodynamics calculations of the internal dynamics of a shaped cross-section tokamak plasma that span the timescales associated with ideal and resistive stability as well as parallel and perpendicular transport. We specify the transport coefficients and apply a ``current controller'' that adjusts the boundary loop-voltage to keep the total plasma current fixed. The 3D 2-fluid plasma model advances the magnetic field, velocities, electron and ion temperatures, and plasma density. We find that the plasma either reaches a stationary quasi-helical state in which the central safety factor is approximately unity, or it periodically undergoes either simple or compound sawtooth oscillations [3] with a period that approaches a constant value. By comparing a dee-shaped cross section with an elliptical shaped cross section, it is shown that the plasma shape has a large effect on determining the sawtooth behavior and the associated mode activity. Application to ITER shaped tokamak plasmas predict the magnitude of the 3D boundary deformation as a result of a stationary quasi-helical state forming in the interior. [4pt] [1] J. Breslau, N. Ferraro, S.C. Jardin, Physics of Plasmas 16 092503 (2009) [0pt] [2] S. C. Jardin, N. Ferraro, J. Breslau, J. Chen, Computational Science and Discovery 5 014002 (2012) [0pt] [3] X. von Goeler, W. Stodiek, and N. Sauthoff, Phys. Rev. Lett. 33, 1201 (1974)
Centre-solenoid-free merging start-up of spherical tokamak plasmas in UTST
NASA Astrophysics Data System (ADS)
Inomoto, M.; Watanabe, T. G.; Gi, K.; Yamasaki, K.; Kamio, S.; Imazawa, R.; Yamada, T.; Guo, X.; Ushiki, T.; Ishikawa, H.; Nakamata, H.; Kawakami, N.; Sugawara, T.; Matsuyama, K.; Noma, K.; Kuwahata, A.; Tanabe, H.
2015-03-01
A centre-solenoid-free merging start-up scheme for spherical tokamak plasmas was developed in a University of Tokyo spherical tokamak (UTST) experiment by using outer poloidal field coils. Torus breakdown was initiated at null points and two spherical tokamak plasmas with a total current up to 80 kA were generated inductively. Their merging process provided substantial ion and electron heating by magnetic reconnection. The obtained dependence of heating on plasma current suggests that high-temperature and high-current plasma suitable for neutral beam injection is attainable under the realistic conditions in the merging start-up method.
Electron temperature gradient driven instability in the tokamak boundary plasma
Xu, X.Q.; Rosenbluth, M.N.; Diamond, P.H.
1992-12-15
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{sup {minus}1/2}e{sup {gamma}mt}.
Reflectometer measurements of density fluctuations in tokamak plasmas
Nazikian, R.; Mazzucato, E.
1994-08-01
We show that many anomalous features observed in reflectometer measurements of turbulent fluctuations in tokamak plasmas, such as loss of coherent reflection, large amplitude fluctuations, large angular divergence of the reflected waves and correlation lengths of the order of the free space wavelength of the probe beam, can be explained by modeling the plasma fluctuations as a poloidally varying random phase grating located at the cutoff with a phase magnitude given by 1D geometric optics. A key result of our analysis is that the turbulence spectrum cannot be inferred from phase measurements when large amplitude fluctuations are observed at the receiver. However, the turbulence spectrum may still be recovered from phase measurements by use of imaging optics, and wide angle phase sensitive receivers.
Turbulent and neoclassical impurity transport in tokamak plasmas
Fueloep, T.; Nordman, H.
2009-03-15
Impurity particle transport in tokamaks is studied using an electrostatic fluid model for main ion and impurity temperature gradient (ITG) mode and trapped electron (TE) mode turbulence in the collisionless limit and neoclassical theory. The impurity flux and impurity density peaking factor obtained from a self-consistent treatment of impurity transport are compared and contrasted with the results of the often used trace impurity approximation. Comparisons between trace and self-consistent turbulent impurity transport are performed for ITER-like profiles. It is shown that for small impurity concentrations the trace impurity limit is adequate if the plasma is dominated by ITG turbulence. However, in case of TE mode dominated plasmas the contribution from impurity modes may be significant, and therefore a self-consistent treatment may be needed.
Turbulent Transport in Tokamak Plasmas with Rotational Shear
Barnes, M.; Highcock, E. G.; Cowley, S. C.; Roach, C. M.
2011-04-29
Nonlinear gyrokinetic simulations are conducted to investigate turbulent transport in tokamak plasmas with rotational shear. At sufficiently large flow shears, linear instabilities are suppressed, but transiently growing modes drive subcritical turbulence whose amplitude increases with flow shear. This leads to a local minimum in the heat flux, indicating an optimal ExB shear value for plasma confinement. Local maxima in the momentum fluxes are observed, implying the possibility of bifurcations in the ExB shear. The critical temperature gradient for the onset of turbulence increases with flow shear at low flow shears; at higher flow shears, the dependence of heat flux on temperature gradient becomes less stiff. The turbulent Prandtl number is found to be largely independent of temperature and flow gradients, with a value close to unity.
Evidence for Self-organized Criticality in Tokamak Plasma Transport
NASA Astrophysics Data System (ADS)
Moyer, R. A.; Lehmer, R.; Rhodes, T. H.; Doyle, E. J.; Peebles, W. A.; Rettig, C. L.; Groebner, R. J.
1998-11-01
Measurements of turbulence spectra and particle flux probability distributions from the DIII-D tokamak exhibit significant agreement with predictions of self organized criticality (SOC) theories. Power spectra of density tilde n, floating potential, and particle flux Γ have three regions of frequency dependence: low frequency f^0, intermediate frequency f-1, and high frequency f-4, consistent with power spectra observed in SOC modeling of various systems. The particle flux probability distribution function P(Γ) for radially outgoing flux shows a Γ-1 dependent region extending over two decades of Γ, a clear indication of self organized behavior. Radially inward flux, representing toppling events up the density gradient (which are outside the scope of the models), also displays a Γ-1 dependent region. These measurements indicate that the plasma is in a state consistent with self organized criticality, and place a significant constraint on plasma transport models.
Numerical simulations of tokamak plasma turbulence and internal transport barriers
NASA Astrophysics Data System (ADS)
Thyagaraja, A.
2000-12-01
A wide variety of magnetically confined plasmas, including many tokamaks such as the JET, TFTR, JT-60U, DIII-D, RTP, show clear evidence for the existence of the so-called `internal transport barriers' (ITBs) which are regions of relatively good confinement, associated with substantial gradients in temperature and/or density. A computational approach to investigating the properties of tokamak plasma turbulence and transport is developed. This approach is based on the evolution of global, two-fluid, nonlinear, electromagnetic plasma equations of motion with specified sources. In this paper, the computational model is applied to the problem of determining the nature and physical characteristics of barrier phenomena, with particular reference to RTP (electron-cyclotron resonance heated) and JET (neutral beam heated) observations of ITBs. The simulations capture features associated with the formation of these ITBs, and qualitatively reproduce some of the observations made on RTP and JET. The picture of plasma turbulence suggested involves variations of temperature and density profiles induced by the electromagnetic fluctuations, on length scales intermediate between the system size and the ion Larmor radius, and time scales intermediate between the confinement time and the Alfvén time (collectively termed `mesoscales'). The back-reaction of such profile `corrugations' (features exhibiting relatively high local spatial gradients and rapid time variations) on the development and saturation of the turbulence itself plays a key role in the nonlinear dynamics of the system. The corrugations are found to modify the dynamical evolution of radial electric field shear and the bootstrap current density, which in turn influence the turbulence. The interaction is mediated by relatively long wavelength, electromagnetic modes excited by an inverse cascade and involving nonlinear instabilities and relaxation phenomena such as intermittency and internal mode locking.
Lee, H.G.; Lee, J.H.; Johnson, D.; Ellis, R.; Feder, R.; Park, H.
2004-10-01
The core and edge Thomson systems on Korea Superconducting Tokamak Advanced Research employ two different sets of lens collection optics. Their collection systems are positioned in the front end of a long reentrant cassette for optimum viewing coverage and optical throughput. Both systems collect the scattered light from a single tangential beam of multiple 50-Hz Nd:YAG lasers and image the scattering volume from core to edge with 40 spatial points. In order to obtain a higher resolution of 5 mm, the edge system has more spatial channels than the core system. Pressure-free heat shield windows, which will absorb the radiation heat flux, are mounted in front of large vacuum windows to protect them from the radiation heat load during long-pulse discharges.
OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM
BURRELL,HK
2002-11-01
OAK A271 OVERVIEW OF RECENT EXPERIMENTAL RESULTS FROM THE DIII-D ADVANCED TOKAMAK PROGRAM. The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last International Atomic Energy Agency (IAEA) meeting, they have made significant progress in developing the building blocks needed for AT operation: (1) they have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, they have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4 {tau}{sub E} limited by the neoclassical tearing mode; (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, they have stabilized the (m,n) = (3,2) neoclassical tearing mode and then increased {beta}{sub T} by 60%; (4) they have produced ECCD stabilization of the (2,1) neoclassical tearing mode in initial experiments; (5) they have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) they have demonstrated stationary tokamak operation for 6.5 s (36 {tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx} 0.4 as ITER but at much higher q{sub 95} = 4.2. The authors have developed general improvements applicable to conventional and advanced tokamak operating modes: (1) they have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, ELM heat load to the divertor and which can run for long periods of time (3.8 s or 25 {tau}{sub E}) with constant density and constant radiated power; (2) they have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet
Overview of recent experimental results from the DIII-D advanced tokamak program.
Burrell, K. H.
2003-12-01
The DIII-D research program is developing the scientific basis for advanced tokamak (AT) modes of operation in order to enhance the attractiveness of the tokamak as an energy producing system. Since the last international atomic energy agency (IAEA) meeting, we have made significant progress in developing the building blocks needed for AT operation: (1) we have doubled the magnetohydrodynamic (MHD) stable tokamak operating space through rotational stabilization of the resistive wall mode; (2) using this rotational stabilization, we have achieved {beta}{sub N}H{sub 89} {ge} 10 for 4{tau}{sub E} limited by the neoclassical tearing mode (NTM); (3) using real-time feedback of the electron cyclotron current drive (ECCD) location, we have stabilized the (m, n) = (3, 2) NTM and then increased {beta}{sub T} by 60%; (4) we have produced ECCD stabilization of the (2, 1) NTM in initial experiments; (5) we have made the first integrated AT demonstration discharges with current profile control using ECCD; (6) ECCD and electron cyclotron heating (ECH) have been used to control the pressure profile in high performance plasmas; and (7) we have demonstrated stationary tokamak operation for 6.5 s (36{tau}{sub E}) at the same fusion gain parameter of {beta}{sub N}H{sub 89}/q{sub 95}{sup 2} {approx_equal} as ITER but at much higher q{sub 95} = 4.2. We have developed general improvements applicable to conventional and AT operating modes: (1) we have an existence proof of a mode of tokamak operation, quiescent H-mode, which has no pulsed, edge localized modes (ELM) heat load to the divertor and which can run for long periods of time (3.8 s or 25{tau}{sub E}) with constant density and constant radiated power; (2) we have demonstrated real-time disruption detection and mitigation for vertical disruption events using high pressure gas jet injection of noble gases; (3) we have found that the heat and particle fluxes to the inner strike points of balanced, double-null divertors are much
CONTROL OF MHD STABILITY IN DIII-D ADVANCED TOKAMAK DISCHARGES
STRAIT,EJ; BIALEK,J; CHANCE,MS; CHU,MS; EDGELL,DH; FERRON,JR; GREENFIELD,CM; GAROFALO,AM; HUMPHREYS,DA; JACKSON,GL; JAYAKUMAR,RJ; JERNIGAN,TC; KIM,JS; LA HAYE,RJ; LAO,LL; LUCE,TC; MAKOWSKI,MA; MURAKAMI,M; NAVRATIL,GA; OKABAYASHI,M; PETTY,CC; REIMERDES,H; SCOVILLE,JT; TURNBULL,AD; WADE,MR; WALKER,ML; WHYTE,DG; DIII-D TEAM
2003-06-01
OAK-B135 Advanced tokamak research in DIII-D seeks to optimize the tokamak approach for fusion energy production, leading to a compact, steady state power source. High power density implies operation at high toroidal beta, {beta}{sub T}=
2{micro}{sub 0}/B{sub T}{sup 2}, since fusion power density increases roughly as the square of the plasma pressure. Steady-state operation with low recirculating power for current drive implies operation at high poloidal beta, {beta}{sub P} =
2{micro}{sub 0}/{sup 2}, in order to maximize the fraction of self-generated bootstrap current. Together, these lead to a requirement of operation at high normalized beta, {beta}{sub N} = {beta}{sub T}(aB/I), since {beta}{sub P}{beta}{sub T} {approx} 25[(1+{kappa}{sup 2})/2] ({beta}{sub N}/100){sup 2}. Plasmas with high normalized beta are likely to operate near one or more stability limits, so control of MHD stability in such plasmas is crucial.
NASA Astrophysics Data System (ADS)
Liu, S. C.; Shao, L. M.; Zweben, S. J.; Xu, G. S.; Guo, H. Y.; Cao, B.; Wang, H. Q.; Wang, L.; Yan, N.; Xia, S. B.; Zhang, W.; Chen, R.; Chen, L.; Ding, S. Y.; Xiong, H.; Zhao, Y.; Wan, B. N.; Gong, X. Z.; Gao, X.
2012-12-01
Gas puff imaging (GPI) offers a direct and effective diagnostic to measure the edge turbulence structure and velocity in the edge plasma, which closely relates to edge transport and instability in tokamaks. A dual GPI diagnostic system has been installed on the low field side on experimental advanced superconducting tokamak (EAST). The two views are up-down symmetric about the midplane and separated by a toroidal angle of 66.6°. A linear manifold with 16 holes apart by 10 mm is used to form helium gas cloud at the 130×130 mm (radial versus poloidal) objective plane. A fast camera is used to capture the light emission from the image plane with a speed up to 390 804 frames/s with 64×64 pixels and an exposure time of 2.156 μs. The spatial resolution of the system is 2 mm at the objective plane. A total amount of 200 Pa.L helium gas is puffed into the plasma edge for each GPI viewing region for about 250 ms. The new GPI diagnostic has been applied on EAST for the first time during the recent experimental campaign under various plasma conditions, including ohmic, L-mode, and type-I, and type-III ELMy H-modes. Some of these initial experimental results are also presented.
High speed cine film studies of plasma behaviour and plasma surface interactions in tokamaks
NASA Astrophysics Data System (ADS)
Goodall, D. H. J.
1982-12-01
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.
Kang, C. S.; Lee, S. G.
2014-07-15
The behavior of relativistic runaway electrons during Electron Cyclotron Resonance Heating (ECRH) discharges is investigated in the Korea Superconducting Tokamak Advanced Research device. The effect of the ECRH on the runaway electron population is discussed. Observations on the generation of superthermal electrons during ECRH will be reported, which will be shown to be consistent with existing theory for the development of a superthermal electron avalanche during ECRH [A. Lazaros, Phys. Plasmas 8, 1263 (2001)].
ADX: a high field, high power density, advanced divertor and RF tokamak
NASA Astrophysics Data System (ADS)
LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.
2015-05-01
The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept
Helical temperature perturbations associated with tearing modes in tokamak plasmas
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.
Chen, Yingjie; Wu, Zhenwei; Gao, Wei; Ti, Ang; Zhang, Ling; Jie, Yinxian; Zhang, Jizong; Huang, Juan; Xu, Zong; Zhao, Junyu
2015-02-01
The multi-channel visible bremsstrahlung measurement system has been developed on Experimental Advanced Superconducting Tokamak (EAST). In addition to providing effective ion charge Zeff as a routine diagnostic, this diagnostic can also be used to estimate other parameters. With the assumption that Zeff can be seen as constant across the radius and does not change significantly during steady state discharges, central electron temperature, averaged electron density, electron density profile, and plasma current density profile have been obtained based on the scaling of Zeff with electron density and the relations between Zeff and these parameters. The estimated results are in good coincidence with measured values, providing an effective and convenient method to estimate other plasma parameters. PMID:25725844
Chen, Yingjie; Wu, Zhenwei; Gao, Wei; Ti, Ang; Zhang, Ling; Jie, Yinxian; Zhang, Jizong; Huang, Juan; Xu, Zong; Zhao, Junyu
2015-02-15
The multi-channel visible bremsstrahlung measurement system has been developed on Experimental Advanced Superconducting Tokamak (EAST). In addition to providing effective ion charge Z{sub eff} as a routine diagnostic, this diagnostic can also be used to estimate other parameters. With the assumption that Z{sub eff} can be seen as constant across the radius and does not change significantly during steady state discharges, central electron temperature, averaged electron density, electron density profile, and plasma current density profile have been obtained based on the scaling of Z{sub eff} with electron density and the relations between Z{sub eff} and these parameters. The estimated results are in good coincidence with measured values, providing an effective and convenient method to estimate other plasma parameters.
Plasma Rotation Under a Driven Radial Current in a Tokamak
NASA Astrophysics Data System (ADS)
Chang, Choong-Seock
1999-11-01
Neoclassical behavior of plasma rotation under a driven radial electrical current is studied in a tokamak geometry. Representative examples of the radial electrical current drive are the fat-banana ion orbit loss to the first wall, resonant particle transport by radio frequency waves, injection of electrons from an emissive probe, and injection of highly anisotropic electrons using ripple transport. An ambipolar radial electric field develops in an MHD time scale in such a way that the driven current is balanced by a return current j^p in the plasma. The initial poloidal rotation, given by E× B, is an immediate transient response of the plasma, which can be very large. The j^p× B torque pushes the plasma into a new rotational equilibrium state both toroidally and poloidally. In general, the initially large poloidal rotation relaxes to a smaller value, but the initially small toroidal rotation increases to a greater value. It is shown that the time scale for the relaxation of poloidal rotation is the same as that of toroidal rotation generation, which is usually given by an anomalous phenomenon.
In situ "artificial plasma" calibration of tokamak magnetic sensors.
Shiraki, D; Levesque, J P; Bialek, J; Byrne, P J; DeBono, B A; Mauel, M E; Maurer, D A; Navratil, G A; Pedersen, T S; Rath, N
2013-06-01
A unique in situ calibration technique has been used to spatially calibrate and characterize the extensive new magnetic diagnostic set and close-fitting conducting wall of the High Beta Tokamak-Extended Pulse (HBT-EP) experiment. A new set of 216 Mirnov coils has recently been installed inside the vacuum chamber of the device for high-resolution measurements of magnetohydrodynamic phenomena including the effects of eddy currents in the nearby conducting wall. The spatial positions of these sensors are calibrated by energizing several large in situ calibration coils in turn, and using measurements of the magnetic fields produced by the various coils to solve for each sensor's position. Since the calibration coils are built near the nominal location of the plasma current centroid, the technique is referred to as an "artificial plasma" calibration. The fitting procedure for the sensor positions is described, and results of the spatial calibration are compared with those based on metrology. The time response of the sensors is compared with the evolution of the artificial plasma current to deduce the eddy current contribution to each signal. This is compared with simulations using the VALEN electromagnetic code, and the modeled copper thickness profiles of the HBT-EP conducting wall are adjusted to better match experimental measurements of the eddy current decay. Finally, the multiple coils of the artificial plasma system are also used to directly calibrate a non-uniformly wound Fourier Rogowski coil on HBT-EP. PMID:23822340
Theory for neoclassical toroidal plasma viscosity in tokamaks
NASA Astrophysics Data System (ADS)
Shaing, K. C.; Chu, M. S.; Hsu, C. T.; Sabbagh, S. A.; Seol, Jae Chun; Sun, Y.
2012-12-01
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.
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.
2015-10-15
Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around k_{θρs} ~ 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Furthermore, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma
NASA Astrophysics Data System (ADS)
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.
2015-10-01
Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E ×B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around kθρs˜0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E ×B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E ×B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Moreover, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport in advanced ST
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.
2015-10-15
Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transportmore » that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around kθρs ~ 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Furthermore, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport in
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.; Li, Z. Q.
2015-10-01
Highly distinct features of spherical tokamaks (ST), such as National Spherical Torus eXperiment (NSTX) and NSTX-U, result in a different fusion plasma regime with unique physics properties compared to conventional tokamaks. Nonlinear global gyrokinetic simulations critical for addressing turbulence and transport physics in the ST regime have led to new insights. The drift wave Kelvin-Helmholtz (KH) instability characterized by intrinsic mode asymmetry is identified in strongly rotating NSTX L-mode plasmas. While the strong E x B shear associated with the rotation leads to a reduction in KH/ion temperature gradient turbulence, the remaining fluctuations can produce a significant ion thermal transport that is comparable to the experimental level in the outer core region (with no "transport shortfall"). The other new, important turbulence source identified in NSTX is the dissipative trapped electron mode (DTEM), which is believed to play little role in conventional tokamak regime. Due to the high fraction of trapped electrons, long wavelength DTEMs peaking around k(theta)rho(s) similar to 0.1 are destabilized in NSTX collisionality regime by electron density and temperature gradients achieved there. Surprisingly, the E x B shear stabilization effect on DTEM is remarkably weak, which makes it a major turbulence source in the ST regime dominant over collisionless TEM (CTEM). The latter, on the other hand, is subject to strong collisional and E x B shear suppression in NSTX. DTEM is shown to produce significant particle, energy and toroidal momentum transport, in agreement with experimental levels in NSTX H-modes. Moreover, DTEM-driven transport in NSTX parametric regime is found to increase with electron collision frequency, providing one possible source for the scaling of confinement time observed in NSTX H-modes. Most interestingly, the existence of a turbulence-free regime in the collision-induced CTEM to DTEM transition, corresponding to a minimum plasma transport
Development of precision measurement network of experimental advanced superconducting tokamak
NASA Astrophysics Data System (ADS)
Yu, Liandong; Zhao, Huining; Zhang, Wei; Li, Weishi; Deng, Huaxia; Song, Yuntao; Gu, Yongqi
2014-12-01
In order to obtain accurate position of the inner key components in the experimental advanced superconducting tokamak (EAST), a combined optical measurement method which is comprised of a laser tracker (LT) and articulated coordinate measuring machine (CMM) has been brought forward. LT, which is an optical measurement instrument and has a large measurement range and high accuracy, is employed for establishing the precision measurement network of EAST, and the articulated CMM is also employed for measuring the inner key components of EAST. The measurement uncertainty analyzed by the Unified Spatial Metrology Network (USMN) is 0.20 mm at a confidence probability of 95.44%. The proposed technology is appropriate for the inspection of the reconstruction of the EAST.
Microwave Doppler reflectometer system in the Experimental Advanced Superconducting Tokamak.
Zhou, C; Liu, A D; Zhang, X H; Hu, J Q; Wang, M Y; Li, H; Lan, T; Xie, J L; Sun, X; Ding, W X; Liu, W D; Yu, C X
2013-10-01
A Doppler reflectometer system has recently been installed in the Experimental Advanced Superconducting (EAST) Tokamak. It includes two separated systems, one for Q-band (33-50 GHz) and the other for V-band (50-75 GHz). The optical system consists of a flat mirror and a parabolic mirror which are optimized to improve the spectral resolution. A synthesizer is used as the source and a 20 MHz single band frequency modulator is used to get a differential frequency for heterodyne detection. Ray tracing simulations are used to calculate the scattering location and the perpendicular wave number. In EAST last experimental campaign, the Doppler shifted signals have been obtained and the radial profiles of the perpendicular propagation velocity during L-mode and H-mode are calculated. PMID:24182112
Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape
Hommen, G.; Baar, M. de; Nuij, P.; Steinbuch, M.; McArdle, G.; Akers, R.
2010-11-15
A new diagnostic is developed to reconstruct the plasma boundary using visible wavelength images. Exploiting the plasma's edge localized and toroidally symmetric emission profile, a new coordinate transform is presented to reconstruct the plasma boundary from a poloidal view image. The plasma boundary reconstruction is implemented in MATLAB and applied to camera images of Mega-Ampere Spherical Tokamak discharges. The optically reconstructed plasma boundaries are compared to magnetic reconstructions from the offline reconstruction code EFIT, showing very good qualitative and quantitative agreement. Average errors are within 2 cm and correlation is high. In the current software implementation, plasma boundary reconstruction from a single image takes 3 ms. The applicability and system requirements of the new optical boundary reconstruction, called OFIT, for use in both feedback control of plasma position and shape and in offline reconstruction tools are discussed.
RF wave propagation and scattering in turbulent tokamak plasmas
Horton, W. Michoski, C.; Peysson, Y.; Decker, J.
2015-12-10
Drift wave turbulence driven by the steep electron and ion temperature gradients in H-mode divertor tokamaks produce scattering of the RF waves used for heating and current drive. The X-ray emission spectra produced by the fast electrons require the turbulence broaden RF wave spectrum. Both the 5 GHz Lower Hybrid waves and the 170 GHz electron cyclotron [EC] RF waves experience scattering and diffraction by the electron density fluctuations. With strong LHCD there are bifurcations in the coupled turbulent transport dynamics giving improved steady-state confinement states. The stochastic scattering of the RF rays makes the prediction of the distribution of the rays and the associated particle heating a statistical problem. Thus, we introduce a Fokker-Planck equation for the probably density of the RF rays. The general frame work of the coupled system of coupled high frequency current driving rays with the low-frequency turbulent transport determines the profiles of the plasma density and temperatures.
Plasma density behavior in the Hefei tokamak-7
NASA Astrophysics Data System (ADS)
Gao, Xiang; Jie, Y. X.; Yang, Y.; Xia, C. Y.; Wei, M. S.; Zhang, S. Y.; Cheng, Y. F.; Hu, L. Q.; Mao, J. S.; Tong, X. D.; Wan, B. N.; Kuang, G. L.; Li, J. G.; Zhao, Y. P.; Luo, J. R.; Qiu, N.; Yang, K.; Li, G.; Xie, J. K.; Wan, Y. X.
2000-07-01
The density profiles were measured in the Hefei tokamak-7 (HT-7) [World Survey of Activities in Controlled Fusion Research, Nuclear Fusion Special Supplement (International Atomic Energy Agency, Vienna, 1997), p. 61] ohmic discharges by means of a new multichannel far-infrared (FIR) laser interferometer. The progress on the extension of the HT-7 ohmic discharge operation region was introduced. The experiment results at the density limit, the multifaceted asymmetric radiation from the edge (MARFE) phenomena, the rf (radio frequency) boronization experiments, and the fueling efficiency studies were reported. The plasma physics in the molecular beam injection (MBI), the pellet injection (PI), and the gas puffing (GP) fueling experiments was studied and discussed.
Continuum kinetic modeling of the tokamak plasma edge
Dorf, M. A.; Dorr, M.; Rognlien, T.; Hittinger, J.; Cohen, R.
2016-03-10
In this study, the first 4D (axisymmetric) high-order continuum gyrokinetic transport simulations that span the magnetic separatrix of a tokamak are presented. The modeling is performed with the COGENT code, which is distinguished by fourth-order finite-volume discretization combined with mapped multiblock grid technology to handle the strong anisotropy of plasmatransport and the complex X-point divertor geometry with high accuracy. The calculations take into account the effects of fully nonlinear Fokker-Plank collisions, electrostatic potential variations, and anomalous radial transport. Topics discussed include: (a) ion orbit loss and the associated toroidal rotation and (b) edge plasma relaxation in the presence of anomalousmore » radial transport.« less
Study of internal transport barrier triggering mechanism in tokamak plasmas
Dong, J.Q.; Mou, Z.Z.; Long, Y.X.; Mahajan, S.M.
2004-12-01
Sheared flow layers driven by magnetic energy, released in tearing-reconnection processes inherent in dissipative magnetohydrodynamics, are proposed as a triggering mechanism for the creation of the internal transport barrier (ITB) in tokamak plasmas. The double tearing mode, mediated by anomalous electron viscosity in configurations with a nonmonotonic safety factor, is investigated as an example. Particular emphasis is placed on the formation of sheared poloidal flow layers in the vicinity of the magnetic islands. A quasilinear simulation demonstrates that the sheared flows induced by the mode have desirable characteristics (lying just outside the magnetic islands), and sufficient levels required for ITB formation. A possible explanation is also proffered for the experimental observation that the transport barriers are preferentially formed in the proximity of low-order rational surfaces.
RF wave propagation and scattering in turbulent tokamak plasmas
NASA Astrophysics Data System (ADS)
Horton, W.; Michoski, C.; Peysson, Y.; Decker, J.
2015-12-01
Drift wave turbulence driven by the steep electron and ion temperature gradients in H-mode divertor tokamaks produce scattering of the RF waves used for heating and current drive. The X-ray emission spectra produced by the fast electrons require the turbulence broaden RF wave spectrum. Both the 5 GHz Lower Hybrid waves and the 170 GHz electron cyclotron [EC] RF waves experience scattering and diffraction by the electron density fluctuations. With strong LHCD there are bifurcations in the coupled turbulent transport dynamics giving improved steady-state confinement states. The stochastic scattering of the RF rays makes the prediction of the distribution of the rays and the associated particle heating a statistical problem. Thus, we introduce a Fokker-Planck equation for the probably density of the RF rays. The general frame work of the coupled system of coupled high frequency current driving rays with the low-frequency turbulent transport determines the profiles of the plasma density and temperatures.
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.
2015-10-30
New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong ExB shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offering one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. This predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.
Chen Yiping; Wang, F. Q.; Hu, L. Q.; Guo, H. Y.; Wu, Z. W.; Zhang, X. D.; Wan, B. N.; Li, J. G.; Zha, X. J.
2013-02-15
In order to actively control power load on the divertor target plates and study the effect of radiative divertor on plasma parameters in divertor plasmas and heat fluxes to the targets, dedicated experiments with Ar impurity seeding have been performed on experimental advanced superconducting tokamak in typical L-mode discharge with single null divertor configuration, ohmic heating power of 0.5 MW, and lower hybrid wave heating power of 1.0 MW. Ar is puffed into the divertor plasma at the outer target plate near the separatrix strike point with the puffing rate 1.26 Multiplication-Sign 10{sup 20} s{sup -1}. The radiative divertor is formed during the Ar puffing. The SOL/divertor plasma in the L-mode discharge with radiative divertor has been modelled by using SOLPS5.2 code package [V. Rozhansky et al., Nucl. Fusion 49, 025007 (2009)]. The modelling shows the cooling of the divertor plasma due to Ar seeding and is compared with the experimental measurement. The changes of peak electron temperature and heat fluxes at the targets with the shot time from the modelling results are similar to the experimental measurement before and during the Ar impurity seeding, but there is a major difference in time scales when Ar affects the plasma in between experiment and modelling.
Drift-tearing magnetic islands in tokamak plasmas
Fitzpatrick, R.; Waelbroeck, F. L.
2008-01-15
A systematic fluid theory of nonlinear magnetic island dynamics in conventional low-{beta}, large aspect-ratio, circular cross-section tokamak plasmas is developed using an extended magnetohydrodynamics model that incorporates diamagnetic flows, ion gyroviscosity, fast parallel electron heat transport, the ion sound wave, the drift wave, and average magnetic field-line curvature. The model excludes the compressible Alfven wave, geodesic field-line curvature, neoclassical effects, and ion Landau damping. A collisional closure is used for plasma dynamics parallel to the magnetic field. Two distinct branches of island solutions are found, namely the 'sonic' and 'hypersonic' branches. Both branches are investigated analytically, using suitable ordering schemes, and in each case the problem is reduced to a relatively simple set of nonlinear differential equations that can be solved numerically via iteration. The solution determines the island phase velocity, relative to the plasma, and the effect of local currents on the island stability. Sonic islands are relatively wide, flatten both the temperature and density profiles, and tend to propagate close to the local ion fluid velocity. Hypersonic islands, on the other hand, are relatively narrow, only flatten the temperature profile, radiate drift-acoustic waves, and tend to propagate close to the local electron fluid velocity. The hypersonic solution branch ceases to exist above a critical island width. Under normal circumstances, both types of island are stabilized by local ion polarization currents.
Li Erzhong; Zhou Ruijie; Hu Liqun
2011-09-15
In the past, the resonant cyclotron interaction between runaway electrons and lower hybrid waves via anomalous Doppler broadening was experimentally investigated, and it was shown to be able to create a barrier to the energy that could be reached by the runaway electrons [E. Li et al., Nucl. Instrum. Methods Phys. Res. A 621, 566 (2010)]. In this paper, to our knowledge for the first time, experimental evidence will be provided for a resonant cyclotron interaction between runaway electrons and magnetohydrodynamics modes in a stochastic magnetic field in the experimental advanced superconducting tokamak (EAST), which has been theoretically proposed as a mechanism able to limit the maximum attainable energy by runaway electrons in tokamak plasmas [J. R. Martin-Solis and R. Sanchez, Phys. Plasmas 15, 112505 (2008)].
The residual zonal flow in tokamak plasmas toroidally rotating at arbitrary velocity
Zhou, Deng
2014-08-15
Zonal flows, initially driven by ion-temperature-gradient turbulence, may evolve due to the neoclassic polarization in a collisionless tokamak plasma. In our previous work [D. Zhou, Nucl. Fusion 54, 042002 (2014)], the residual zonal flow in a tokamak plasma rotating toroidally at sonic speed is found to have the same form as that of a static plasma. In the present work, the form of the residual zonal flow is presented for tokamak plasmas rotating toroidally at arbitrary velocity. The gyro-kinetic equation is analytically solved for low speed rotation to give the expression of residual zonal flows, and the expression is then generalized for cases with arbitrary rotating velocity through interpolation. The zonal flow level decreases as the rotating velocity increases. The numerical evaluation is in good agreement with the former simulation result for high aspect ratio tokamaks.
3D simulation studies of tokamak plasmas using MHD and extended-MHD models
Park, W.; Chang, Z.; Fredrickson, E.; Fu, G.Y.
1996-12-31
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-{beta} disruption studies in reversed shear plasmas using the MHD level MH3D code, {omega}{sub *i} stabilization and nonlinear island saturation of TAE mode using the hybrid particle/MHD level MH3D-K code, and unstructured mesh MH3D{sup ++} code studies. In particular, three internal mode disruption mechanisms are identified from simulation results which agree which agree well with experimental data.
Transport bifurcation induced by sheared toroidal flow in tokamak plasmas
Highcock, E. G.; Barnes, M.; Roach, C. M.; Cowley, S. C.
2011-10-15
First-principles numerical simulations are used to describe a transport bifurcation in a differentially rotating tokamak plasma. Such a bifurcation is more probable in a region of zero magnetic shear than one of finite magnetic shear, because in the former case the component of the sheared toroidal flow that is perpendicular to the magnetic field has the strongest suppressing effect on the turbulence. In the zero-magnetic-shear regime, there are no growing linear eigenmodes at any finite value of flow shear. However, subcritical turbulence can be sustained, owing to the existence of modes, driven by the ion temperature gradient and the parallel velocity gradient, which grow transiently. Nonetheless, in a parameter space containing a wide range of temperature gradients and velocity shears, there is a sizeable window where all turbulence is suppressed. Combined with the relatively low transport of momentum by collisional (neoclassical) mechanisms, this produces the conditions for a bifurcation from low to high temperature and velocity gradients. A parametric model is constructed which accurately describes the combined effect of the temperature gradient and the flow gradient over a wide range of their values. Using this parametric model, it is shown that in the reduced-transport state, heat is transported almost neoclassically, while momentum transport is dominated by subcritical parallel-velocity-gradient-driven turbulence. It is further shown that for any given input of torque, there is an optimum input of heat which maximises the temperature gradient. The parametric model describes both the behaviour of the subcritical turbulence (which cannot be modelled by the quasi-linear methods used in current transport codes) and the complicated effect of the flow shear on the transport stiffness. It may prove useful for transport modelling of tokamaks with sheared flows.
Kinetic modelling of runaway electron avalanches in tokamak plasmas
NASA Astrophysics Data System (ADS)
Nilsson, E.; Decker, J.; Peysson, Y.; Granetz, R. S.; Saint-Laurent, F.; Vlainic, M.
2015-09-01
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, the
Linear Analysis of Drift Ballooning Modes in Tokamak Edge Plasmas
NASA Astrophysics Data System (ADS)
Tangri, Varun; Kritz, Arnold; Rafiq, Tariq; Pankin, Alexei
2012-10-01
The H-mode pedestal structure depends on the linear stability of drift ballooning modes (DBMs) in many H-mode pedestal models. Integrated modeling that uses these pedestal models requires fast evaluation of linear stability of DBMs. Linear analysis of DBMs is also needed in the computations of effective diffusivities associated with anomalous transport that is driven by the DBMs in tokamak edge plasmas. In this study several numerical techniques of linear analysis of the DBMs are investigated. Differentiation matrix based spectral methods are used to compute the physical eigenvalues of the DBMs. The model for DBMs used here consists of six differential equations [T. Rafiq et al. Phys. Plasmas, 17, 082511, (2010)]. It is important to differentiate among non-physical (numerical) modes and physical modes. The determination of the number of eigenvalues is solved by a computation of the `nearest' and `ordinal' distances. The Finite Difference, Hermite and Sinc based differentiation matrices are used. It is shown that spectral collocation methods are more accurate than finite difference methods. The technique that has been developed for calculating eigenvalues is quite general and is relevant in the computation of other modes that utilize the ballooning mode formalism.
Nonlinear transport processes in tokamak plasmas. I. The collisional regimes
Sonnino, Giorgio; Peeters, Philippe
2008-06-15
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{sup 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{sup 2} when the nonlinear contributions are duly taken into account but, there is still a factor of 10{sup 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.
Liu, D. M. Zhao, W. Z.; He, Y. G.; Chen, B.; Wan, B. N.; Shen, B.; Huang, J.; Liu, H. Q.
2014-11-15
A high-performance integrator is one of the key electronic devices for reliably controlling plasma in the experimental advanced superconducting tokamak for long pulse operation. We once designed an integrator system of real-time drift compensation, which has a low integration drift. However, it is not feasible for really continuous operations due to capacitive leakage error and nonlinearity error. To solve the above-mentioned problems, this paper presents a new alternating integrator. In the new integrator, the integrator system of real-time drift compensation is adopted as one integral cell while two such integral cells work alternately. To achieve the alternate function, a Field Programmable Gate Array built in the digitizer is utilized. The performance test shows that the developed integrator with the integration time constant of 20 ms has a low integration drift (<15 mV) for 1000 s.
Liu, D M; Wan, B N; Zhao, W Z; Shen, B; He, Y G; Chen, B; Huang, J; Liu, H Q
2014-11-01
A high-performance integrator is one of the key electronic devices for reliably controlling plasma in the experimental advanced superconducting tokamak for long pulse operation. We once designed an integrator system of real-time drift compensation, which has a low integration drift. However, it is not feasible for really continuous operations due to capacitive leakage error and nonlinearity error. To solve the above-mentioned problems, this paper presents a new alternating integrator. In the new integrator, the integrator system of real-time drift compensation is adopted as one integral cell while two such integral cells work alternately. To achieve the alternate function, a Field Programmable Gate Array built in the digitizer is utilized. The performance test shows that the developed integrator with the integration time constant of 20 ms has a low integration drift (<15 mV) for 1000 s. PMID:25430391
Analog integrator for the Korea superconducting tokamak advanced research magnetic diagnostics
NASA Astrophysics Data System (ADS)
Bak, J. G.; Lee, S. G.; Son, D.; Ga, E. M.
2007-04-01
An analog integrator, which automatically compensates an integrating drift, has been developed for the magnetic diagnostics in the Korea superconducting tokamak advanced research (KSTAR). The compensation of the drift is done by the analog to digital converter-register-digital to analog converter in the integrator. The integrator will be used in the equilibrium magnetic field measurements by using inductive magnetic sensors during a plasma discharge in the KSTAR machine. Two differential amplifiers are added to the signal path between each magnetic sensor and the integrator in order to improve the performance of the integrator because a long signal cable of 100 m will be used for the measurement in the KSTAR machine. In this work, the characteristics of the integrator with two differential amplifiers are experimentally investigated.
ADVANCES IN COMPREHENSIVE GYROKINETIC SIMULATIONS OF TRANSPORT IN TOKAMAKS
WALTZ RE; CANDY J; HINTON FL; ESTRADA-MILA C; KINSEY JE
2004-10-01
A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate core turbulent transport in actual experimental profiles and enable direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite {beta}, equilibrium ExB shear stabilization, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius ({rho}{sub *}) so as to treat the profile shear stabilization and nonlocal effects which can break gyroBohm scaling. The code operates in either a cyclic flux-tube limit (which allows only gyroBohm scaling) or a globally with physical profile variation. Rohm scaling of DIII-D L-mode has been simulated with power flows matching experiment within error bars on the ion temperature gradient. Mechanisms for broken gyroBohm scaling, neoclassical ion flows embedded in turbulence, turbulent dynamos and profile corrugations, plasma pinches and impurity flow, and simulations at fixed flow rather than fixed gradient are illustrated and discussed.
Protection of tokamak plasma facing components by a capillary porous system with lithium
NASA Astrophysics Data System (ADS)
Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.
2015-08-01
Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.
Lampert, M; Anda, G; Czopf, A; Erdei, G; Guszejnov, D; Kovácsik, Á; Pokol, G I; Réfy, D; Nam, Y U; Zoletnik, S
2015-07-01
A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties. PMID:26233377
Lampert, M.; Anda, G.; Réfy, D.; Zoletnik, S.; Czopf, A.; Erdei, G.; Guszejnov, D.; Kovácsik, Á.; Pokol, G. I.; Nam, Y. U.
2015-07-15
A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera’s measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.
Yu Yaowei; Kim, Young-Ok; Kim, Hak-Kun; Kim, Hong-Tack; Kim, Woong-Chae; Kim, Kwang-Pyo; Son, Soo-Hyun; Bang, Eun-Nam; Hong, Suk-Ho; Yoon, Si-Woo; Zhuang Huidong; Chen Zhongyong
2012-12-15
Massive gas injection (MGI) system was developed on Korea Superconducting Tokamak Advanced Research (KSTAR) in 2011 campaign for disruption studies. The MGI valve has a volume of 80 ml and maximum injection pressure of 50 bar, the diameter of valve orifice to vacuum vessel is 18.4 mm, the distance between MGI valve and plasma edge is {approx}3.4 m. The MGI power supply employs a large capacitor of 1 mF with the maximum voltage of 3 kV, the valve can be opened in less than 0.1 ms, and the amount of MGI can be controlled by the imposed voltage. During KSTAR 2011 campaign, MGI disruptions are carried out by triggering MGI during the flat top of circular and limiter discharges with plasma current 400 kA and magnetic field 2-3.5 T, deuterium injection pressure 39.7 bar, and imposed voltage 1.1-1.4 kV. The results show that MGI could mitigate the heat load and prevent runaway electrons with proper MGI amount, and MGI penetration is deeper under higher amount of MGI or lower magnetic field. However, plasma start-up is difficult after some of D{sub 2} MGI disruptions due to the high deuterium retention and consequently strong outgassing of deuterium in next shot, special effort should be made to get successful plasma start-up after deuterium MGI under the graphite first wall.
Plasma Position Measurements in a Tokamak with an Iron Core Transformer
NASA Astrophysics Data System (ADS)
Kwon, Gi-Chung; Choe, W.; Kim, Jayhyun; Yi, Hyo-Suk; Jeon, Sang-Jean; Huh, Songwhe; Chang, Hong-Young; Choi, Duk-In
2000-07-01
Two simple methods of estimating the plasma position in a large-aspect-ratio, low-βp tokamak with an iron core transformer are demonstrated: a magnetic diagnostic method and an optical method. The magnetic diagnostic method utilizes an array of magnetic pickup coils to measure the poloidal magnetic field produced by the plasma current. To include the effects of toroidicity and an iron core transformer, the correction factor was calculated with the magnetic material (or iron core) inside the calculation domain and incorporated in the analysis. The evolution of horizontal and vertical displacement of the plasma center obtained in this way is used to control the KAIST-Tokamak plasmas. To compare the plasma position estimated using the magnetic pickup coils, a simple optical method is also demonstrated on KAIST-TOKAMAK using a composite video signal from a charge-coupled device (CCD) camera. The two results are in good agreement.
1994-05-27
If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).
High-Q plasmas in the TFTR tokamak
NASA Astrophysics Data System (ADS)
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.-L.; Young, K. M.; Zarnstorff, M. C.; Zweben, S. J.
1991-08-01
In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength Sn and D-D fusion power gain QDD are realized in the neutral-beam-fueled and heated ``supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, Sn increases approximately as P1.8b. The highest-Q shots are characterized by high Te (up to 12 keV), Ti (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad Te profiles, and lower Zeff. 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, QDD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [ne(0)/
Stabilization of Tokamak Plasmas by the Addition of Nonaxisymmetric Coils
NASA Astrophysics Data System (ADS)
Reiman, Allan
2008-11-01
It has been recognized since the early days of the fusion program that stellarator coils can be used to stabilize current carrying, toroidal, magnetically confined plasmas.[1] More recently, it has been shown that the vertical mode in a tokamak can be stabilized by a relatively simple set of parallelogram-shaped, localized, nonaxisymmetric coils.[2] We show that by superposing sets of these parallelogram-shaped, nonaxisymmetric coils at different locations, it is possible to reproduce the coil current patterns for conventional stellarator coils as well as those for Furth-Hartman coils[3]. This allows us to gain insight into the physics of stabilization produced by various sets of nonaxisymmetric coils by analysis of the effect on stability of localized coils at differing locations. In particular, the relationship between the stabilization effect and the rotational transform generated by the nonaxisymmetric coils is clarified. [1] J. L. Johnson, C. R. Oberman, R. M. Kulsrud, and E. A. Frieman, Phys. Fluids 1, 281 (1958) [2] A. Reiman, Phys. Rev. Lett. 99, 135007, (2007). [3] H.P. Furth and C.W. Hartman, Phys. Fluids 11, 408 (1968).
Impurity effects on trapped electron mode in tokamak plasmas
NASA Astrophysics Data System (ADS)
Du, Huarong; Wang, Zheng-Xiong; Dong, J. Q.
2016-07-01
The effects of impurity ions on the trapped electron mode (TEM) in tokamak plasmas are numerically investigated with the gyrokinetic integral eigenmode equation. It is shown that in the case of large electron temperature gradient ( η e ), the impurity ions have stabilizing effects on the TEM, regardless of peaking directions of their density profiles for all normalized electron density gradient R / L n e . Here, R is the major radius and L n e is the electron density gradient scale length. In the case of intermediate and/or small η e , the light impurity ions with conventional inwardly (outwardly) peaked density profiles have stabilizing effects on the TEM for large (small) R / L n e , while the light impurity ions with steep inwardly (outwardly) peaked density profiles can destabilize the TEM for small (large) R / L n e . Besides, the TEM driven by density gradient is stabilized (destabilized) by the light carbon or oxygen ions with inwardly (outwardly) peaked density profiles. In particular, for flat and/or moderate R / L n e , two independent unstable modes, corresponding respectively to the TEM and impurity mode, are found to coexist in plasmas with impurity ions of outwardly peaked density profiles. The high Z tungsten impurity ions play a stronger stabilizing role in the TEM than the low Z impurity ions (such as carbon and oxygen) do. In addition, the effects of magnetic shear and collision on the TEM instability are analyzed. It is shown that the collisionality considered in this work weakens the trapped electron response, leading to a more stable TEM instability, and that the stabilizing effects of the negative magnetic shear on the TEM are more significant when the impurity ions with outwardly peaked density profile are taken into account.
Dynamics and Feedback Control of Plasma Equilibrium Position in a Tokamak.
NASA Astrophysics Data System (ADS)
Burenko, Oleg
A brief history of the beginnings of nuclear fusion research involving toroidal closed-system magnetic plasma containment is presented. A tokamak machine is defined mathematically for the purposes of plasma equilibrium position perturbation analysis. The perturbation equations of a tokamak plasma equilibrium position are developed. Solution of the approximated perturbation equations is carried out. A unique, simple, and useful plasma displacement dynamics transfer function of a tokamak is developed. The dominant time constants of the dynamics transfer function are determined in a symbolic form. This symbolic form of the dynamics transfer function makes it possible to study the stability of a tokamak's plasma equilibrium position. Knowledge of the dynamics transfer function permits systematic syntheses of the required plasma displacement feedback control systems. The major parameters governing the plasma equilibrium position stability of a tokamak are shown to be (1) external magnetic field decay index, (2) transformer iron core effect, (3) plasma current, (4) radial rate-of-change inductance parameter, (5) vertical rate-of-change inductance parameter, and (6) vacuum vessel eddy-current time constant. An important and unique result is derived, showing that for a vacuum vessel eddy-current time constant exceeding a certain value the vertical plasma equilibrium position is stable, in spite of an intentional vertical instability design represented by a negative decay index. It is shown that a tokamak design having a theoretical set of positive decay index, negative radical rate-of-change inductance parameter, and positive vertical rate-of-change inductance parameter is expected to have a better plasma equilibrium position stability tolerance than a tokamak design having the same set with the signs reversed. The results of an actual hardware ISX-A tokamak plasma displacement feed-back control system design are presented. It is shown that a theoretical design computer
Spectroscopy of smooth deuterated carbon films redeposited from plasma discharge in the tokamak T-10
Svechnikov, N. Yu. Stankevich, V. G.; Lebedev, A. M.; Men'shikov, K. A.; Kolbasov, B. N.; Kriventsov, V. V.
2006-12-15
Smooth deuterated carbon films redeposited from a deuterium plasma discharge in the tokamak T-10 vacuum chamber have been investigated by different spectroscopic methods and temperature measurements. The photoluminescence excitation spectra of sp{sup 3}-sp{sup 2} nanostructures of tokamak films and sp{sup 2} nanostructures of fullerite C60 films are compared. The effect of defect states on the photoluminescence and its temperature quenching is discussed. It is concluded that the mechanism of thermal luminescence quenching for smooth deuterated tokamak films is close to the corresponding mechanism for amorphous a-C:H films.
High-Q plasmas in the TFTR tokamak
Jassby, D.L.; Bell, M.G.; Bitter, M.; Boivin, R.; Bretz, N.L.; Budny, R.V.; 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.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Mueller, D.; Owens, D
1991-05-01
In the Tokamak Fusion Test Reactor, the highest neutron source strength S{sub n} and D-D fusion power gain Q{sub DD} are realized in the neutral-beam fueled and heated supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, S{sub n} increases approximately as P{sub b}{sup 1.8}. The highest-Q shots are characterized by high T{sub e}, T{sub i}, and stored energy highly peaked density profiles, broad T{sub e} profiles, and lower Z{sub eff}. Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles, and improved alignment with the plasma, have mitigated the carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, Q{sub DD} increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness during the beam pulse. To date the best fusion results are S{sub n} = 5 {times} 10{sup 16} n/s, Q{sub DD} = 1.85 {times} 10{sup {minus}3}, and neutron yield = 4.0 {times} 10{sup 16} n/pulse, obtained at I{sub p} = 1.6 to 1.9 MA and beam energy E{sub b} = 95 to 103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50--60% of S{sub n} arises from beam-target reactions, with the remainder divided between beam-beam and thermonuclear reactions, the thermonuclear fraction increasing with P{sub b}. The simulations predict that Q{sub DT} = 0.3 to 0.4 would be obtained for the best present plasma conditions, if half the deuterium neutral beams were to be replaced by tritium beams. Somewhat higher values are calculated if D beams are injected into a predominantly tritium target plasma. The projected central beta of fusion alphas is 0.4--0.6%, a level sufficient for the study of alpha-induced collective effects. 16 refs., 8 figs., 3 tabs.
NASA Astrophysics Data System (ADS)
Ivanov, A. A.; Martynov, A. A.; Medvedev, S. Yu.; Poshekhonov, Yu. Yu.
2015-03-01
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.
Ivanov, A. A. Martynov, A. A. Medvedev, S. Yu. Poshekhonov, Yu. Yu.
2015-03-15
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.
Poloidal rotation near the edge of a tokamak plasma in [ital H] mode
Hinton, F.L.; Kim, J.; Kim, Y.; Brizard, A.; Burrell, K.H. )
1994-02-21
Ion poloidal flow in tokamaks near the plasma edge has been calculated by extending neoclassical theory to include orbit squeezing, which is the reduction of the ion banana widths due to the gradient in the radial electric field. The calculated poloidal flow velocity is a significant fraction of the ion diamagnetic velocity, which can be much larger than the velocity predicted by neoclassical theory (proportional to the ion temperature gradient). The agreement with spectroscopic measurements of the poloidal rotation velocity in helium plasmas in the DIII-D tokamak is shown to be reasonably good very close to the plasma edge.
Generation of plasma rotation by ion cyclotron resonance heating in tokamaks
Chang, C.S.; Phillips, C.K.; White, R.; Zweben, S.; Bonoli, P.T.; Rice, J.E.; Greenwald, M.J.; deGrassie, J.
1999-05-01
A physical mechanism for generation of a plasma rotation and radial electric field by ion cyclotron resonance heating (ICRH) is presented in a tokamak geometry. By breaking the omnigenity of resonant ion orbits, ICRH can induce a nonambipolar minor-radial transport of resonant ions. This yields a radial charge separation, a modification to radial electric field E{sub r}, and the generation of plasma rotation. It is estimated that the ICRH fast-wave power available in the present-day tokamak experiments can be large enough to give a significant modification to plasma rotation. {copyright} {ital 1999 American Institute of Physics.}
Severo, J. H. F.; Nascimento, I. C.; Kuznetov, Yu. K.; Tsypin, V. S.; Galvao, R. M. O.; Tendler, M.
2007-04-15
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.
First results on fast wave current drive in advanced tokamak discharges in DIII-D
Prater, R.; Cary, W.P.; Baity, F.W.
1995-07-01
Initial experiments have been performed on the DIII-D tokamak on coupling, direct electron heating, and current drive by fast waves in advanced tokamak discharges. These experiments showed efficient central heating and current drive in agreement with theory in magnitude and profile. Extrapolating these results to temperature characteristic of a power plant (25 keV) gives current drive efficiency of about 0.3 MA/m{sup 2}.
Analysis of pedestal gradient characteristic on the Experimental Advanced Superconducting Tokamak
NASA Astrophysics Data System (ADS)
Wang, Teng Fei; Han, Xiao Feng; Zang, Qing; Xiao, Shu Mei; Tian, Bao Gang; Hu, Ai Lan; Zhao, Jun Yu
2016-05-01
A pedestal database was built based on type I edge localized mode H-modes in the Experimental Advanced Superconducting Tokamak. The most common functional form hyperbolic tangent function (tanh) method is used to analyze pedestal characteristics. The pedestal gradient scales linearly with its pedestal top and the normalized pedestal pressure gradient α shows a strong correlation with electron collisionality. The connection among pedestal top value, gradient, and width is established with the normalized pedestal pressure gradient. In the core region of the plasma, the nature of the electron temperature stiffness reflects a proportionality between core and pedestal temperature while the increase proportion is lower than that expected in the high temperature region. However, temperature profile stiffness is limited or even disappears at the edge of the plasma, while the gradient length ratio ( ηe ) on the pedestal is important. The range of ηe is from 0.5 to 2, varying with the plasma parameters. The pedestal temperature brings a more significant impact on ηe than pedestal density.
NASA Astrophysics Data System (ADS)
Guo, H. Y.; Li, J.; Wan, B. N.; Gong, X. Z.; Liang, Y. F.; Xu, G. S.; Zhang, X. D.; Ding, S. Y.; Gan, K. F.; Hu, J. S.; Hu, L. Q.; Liu, S. C.; Qian, J. P.; Sun, Y. W.; Wang, H. Q.; Wang, L.; Xia, T. Y.; Xiao, B. J.; Zeng, L.; Zhao, Y. P.; Denner, P.; Ferron, J. R.; Garofalo, A. M.; Holcomb, C. T.; Hyatt, A. W.; Jackson, G. L.; Loarte, A.; Maingi, R.; Menard, J. E.; Rack, M.; Solomon, W. M.; Xu, X. Q.; Van Zeeland, M.; Zou, X. L.
2014-05-01
A long-pulse high confinement plasma regime known as H-mode is achieved in the Experimental Advanced Superconducting Tokamak (EAST) with a record duration over 30 s, sustained by Lower Hybrid wave Current Drive (LHCD) with advanced lithium wall conditioning and divertor pumping. This long-pulse H-mode plasma regime is characterized by the co-existence of a small Magneto-Hydrodynamic (MHD) instability, i.e., Edge Localized Modes (ELMs) and a continuous quasi-coherent MHD mode at the edge. We find that LHCD provides an intrinsic boundary control for ELMs, leading to a dramatic reduction in the transient power load on the vessel wall, compared to the standard Type I ELMs. LHCD also induces edge plasma ergodization, broadening heat deposition footprints, and the heat transport caused by ergodization can be actively controlled by regulating edge plasma conditions, thus providing a new means for stationary heat flux control. In addition, advanced tokamak scenarios have been newly developed for high-performance long-pulse plasma operations in the next EAST experimental campaign.
Linear optimal control of tokamak fusion devices
Kessel, C.E.; Firestone, M.A.; Conn, R.W.
1989-05-01
The control of plasma position, shape and current in a tokamak fusion reactor is examined using linear optimal control. These advanced tokamaks are characterized by non up-down symmetric coils and structure, thick structure surrounding the plasma, eddy currents, shaped plasmas, superconducting coils, vertically unstable plasmas, and hybrid function coils providing ohmic heating, vertical field, radial field, and shaping field. Models of the electromagnetic environment in a tokamak are derived and used to construct control gains that are tested in nonlinear simulations with initial perturbations. The issues of applying linear optimal control to advanced tokamaks are addressed, including complex equilibrium control, choice of cost functional weights, the coil voltage limit, discrete control, and order reduction. Results indicate that the linear optimal control is a feasible technique for controlling advanced tokamaks where the more common classical control will be severely strained or will not work. 28 refs., 13 figs.
Extended neoclassical transport theory for incompressible tokamak plasmas
Shaing, K.C.
1997-09-01
Conventional neoclassical transport theory is extended to include the effects of orbit squeezing, and to allow the effective poloidal Mach number U{sub pM}=[(V{sub {parallel}}/v{sub t})+(V{sub E}B/v{sub t}B{sub p})] of the order of unity for incompressible tokamak plasmas. Here, V{sub {parallel}} is the parallel mass flow, v{sub t} is the ion thermal speed, V{sub E} is the poloidal {bold E{times}B} drift speed, B is the magnetic field strength, and B{sub p} is the poloidal magnetic field strength. It is found that ion thermal conductivity is reduced from its conventional neoclassical value in both banana and plateau regimes if U{sub pM}{gt}1 and S{gt}1. Here, S=[1+cI{sup 2}{Phi}{sup {prime}{prime}}/({Omega}{sub 0}B{sub 0})] is the orbit squeezing factor with c the speed of light, I=RB{sub t}, R the major radius, {Phi} the electrostatic potential, B{sub 0} the magnetic field strength on the axis, {Omega}{sub 0}=eB{sub 0}/Mc, M the ion mass, e the ion charge, {Phi}{sup {prime}{prime}}=d{sup 2}{Phi}/d{psi}{sup 2}, and {psi} the poloidal flux function. However, there is an irreducible minimum for the ion thermal conductivity in the banana-plateau regime set by the conventional Pfirsch{endash}Schl{umlt u}ter transport. {copyright} {ital 1997 American Institute of Physics.}
Short wavelength trapped electron modes in tokamak plasmas
NASA Astrophysics Data System (ADS)
Zhang, N.; Gong, X. Y.; Dong, J. Q.; Huang, Q. H.; Gong, L.; Li, J. C.
2016-04-01
The collisionless trapped electron modes in the short wavelength region k⊥ρs>1 (SWTEMs) are studied with the gyrokinetic integral eigenmode equation in tokamak plasmas. Here, we present a systematic study of the correlation between the SWTEMs and short wavelength ion temperature gradient (SWITG) modes. The kθρs spectra of TEM have double humps in the short wavelength and long wavelength regions, respectively. The SWITG modes with trapped electron effects taking into account have broader kθρs spectra. Dependences of growth rate and real frequency of SWTEMs on the various parameters, such as ion temperature gradient (ηi), the temperature gradient of trapped electrons (ηe), toroidicity (ɛn), magnetic shear ( s ̂ ), safety factor (q), and the ratio of temperature (Te/Ti), are investigated in detail. It is found that the SWTEMs propagate in the electron diamagnetic drift direction and require temperature gradient of trapped electrons ηe exceeding thresholds. Moreover, the ion temperature gradient has a strong stabilizing effect on the SWTEMs. The SWTEMs become stable in both regimes of toroidicity ɛn > 0.1 and magnetic shear s ̂>0.5 regardless of the fraction of trapped electrons. In addition, the properties of short wavelength ITG (SWITG) modes are discussed with different ratio of trapped electrons. It is found that trapped electrons of greater fraction have a stronger destabilizing effect on the SWTEM and SWITG modes. These results are significant for the electrons anomalous transport experiments in the future.
Impedance of an intense plasma-cathode electron source for tokamak startup
NASA Astrophysics Data System (ADS)
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.
Impedance of an intense plasma-cathode electron source for tokamak startup
Hinson, Edward Thomas; Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus Galen; Fonck, Raymond J.; Perry, Justin M.
2016-05-31
In this study, an impedance model is formulated and tested for the ~1kV, ~1kA/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 the injected current, Iinj, increases with Vinj according to the standard double layer scaling Iinj ~ Vinj3/2 at low current and transitions to Iinj ~ Vinj1/2more » 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.« less
Kinetic shear Alfvén instability in the presence of impurity ions in tokamak plasmas
Lu, Gaimin; Shen, Y.; Xie, T.; He, Zhixiong; He, Hongda; Qi, Longyu; Cui, Shaoyan
2013-10-15
The effects of impurity ions on the kinetic shear Alfvén (KSA) instability in tokamak plasmas are investigated by numerically solving the integral equations for the KSA eigenmode in the toroidal geometry. The kinetic effects of hydrogen and impurity ions, including transit motion, finite ion Larmor radius, and finite-orbit-width, are taken into account. Toroidicity induced linear mode coupling is included through the ballooning-mode representation. Here, the effects of carbon, oxygen, and tungsten ions on the KSA instability in toroidal plasmas are investigated. It is found that, depending on the concentration and density profile of the impurity ions, the latter can be either stabilizing or destabilizing for the KSA modes. The results here confirm the importance of impurity ions in tokamak experiments and should be useful for analyzing experimental data as well as for understanding anomalous transport and control of tokamak plasmas.
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.
Advanced plasma diagnostics for plasma processing
NASA Astrophysics Data System (ADS)
Malyshev, Mikhail Victorovich
1999-10-01
A new, non-intrusive, non-perturbing diagnostic method was developed that can be broadly applied to low pressure, weakly ionized plasmas and glow discharges-trace rare gases optical emission spectroscopy (TRG-OES). The method is based on a comparison of intensities of atomic emission from trace amounts of inert gases (He, Ne, Ar, Kr, and Xe) that are added to the discharge to intensities calculated from the theoretical model. The model assumes a Maxwellian electron energy distribution function (EEDF), computes the population of emitting levels both from the ground state and the metastable states of rare gases, and from the best fit between theory and experiment determines electron temperature (Te). Subject to conditions, TRG-OES can also yield electron density or its upper or lower limit. From the comparison of the emission from levels excited predominantly by high energy electrons to that excited by low energy electrons, information about the EEDF can be obtained. The use of TRG-OES also allows a traditionally qualitative actinometry technique (determination of concentration of radical species in plasma through optical emission) to become a precise quantitative method by including Te and rare gases metastables effects. A combination of TRG-OES, advanced actinometry, and Langmuir probe measurements was applied to several different plasma reactors and regimes of operation. Te measurements and experiments to correct excitation cross section were conducted in a laboratory helical resonator. Two chamber configuration of a commercial (Lam Research) metal etcher were studied to determine the effects of plasma parameters on plasma-induced damage. Two different methods (RF inductive coupling and ultra-high frequency coupling) for generating a plasma in a prototype reactor were also studied. Pulsed plasmas, a potential candidate to eliminate the plasma-induced damage to microelectronics devices that occurs in manufacturing due to differential charging of the wafer, have
Main Physical Factors Limiting the Accuracy of Polarimetric Measurements in Tokamak Plasma
NASA Astrophysics Data System (ADS)
Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco
The paper reviews and discusses the main factors, limiting the accuracy of polarimetric measurements in tokamak plasma. Theoretical methods, describing evolution of polarimetry state in tokamak plasma, are demonstrated not to contribute noticeably to inaccuracy at sufficiently short beam wavelengths. Based on the literature data as well as on our preliminary estimates it is possible to conclude that the following factors dominate: i) calibration procedure; ii) refraction in the inhomogeneous plasma; iii) influence of weak relativistic effects on plasma dielectric permittivity. The contribution of these factors to is within the range of several per cent. Other causes of measurement inaccuracies (absorption in plasma, diffraction of sounding beam, ray torsion, nonstationary processes in plasma) seem to be less significant.
Integrated Plasma Simulation of Lower Hybrid Current Drive in Tokamaks
NASA Astrophysics Data System (ADS)
Bonoli, P. T.; Wright, J. C.; Harvey, R. W.; Batchelor, D. B.; Berry, L. A.; Kessel, C. E.; Jardin, S. C.
2012-03-01
It has been shown in Alcator C-Mod that the onset time for sawteeth can be delayed significantly (up to 0.5 s) relative to ohmically heated plasmas, through the injection of off-axis LH current drive power [1]. We are simulating these experiments using the Integrated Plasma Simulator (IPS) [2], where the driven LH current density profiles are computed using a ray tracing component (GENRAY) and Fokker Planck code (CQL3D) [3] that are run in a tightly coupled time advance. The background plasma is evolved using the TSC transport code with the Porcelli sawtooth model [4]. Predictions of the driven LH current profiles will be compared with simpler ``reduced'' models for LHCD such as the LSC code which is implemented in TSC and which is also invoked within the IPS. [4pt] [1] C. E. Kessel et al, Bull. of the Am. Phys. Soc. 53, Poster PP6.00074 (2008). [0pt] [2] D. Batchelor et al, Journal of Physics: Conf. Series 125, 012039 (2008). [0pt] [3] R. W. Harvey and M. G. McCoy, Proc. of the IAEA Tech. Comm. Meeting on Simulation and Modeling of Therm. Plasmas, Montreal, Canada (1992). [0pt] [4] S. C. Jardin et al, J. Comp. Phys. 66, 481 (1986).
Texas Experimental Tokamak, a plasma research facility: Technical progress report
Wootton, A.J.
1995-08-01
In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively.
An Advanced Tokamak Fusion Nuclear Science Facility (FNSF-AT)
NASA Astrophysics Data System (ADS)
Chan, V. S.; Garofalo, A. M.; Stambaugh, R. D.
2010-11-01
A Fusion Development Facility (FDF) is a candidate for FNSF-AT. It is a compact steady-state machine of moderate gain that uses AT physics to provide the neutron fluence required for fusion nuclear science development. FDF is conceived as a double-null plasma with high elongation and triangularity, predicted to allow good confinement of high plasma pressure. Steady-state is achieved with high bootstrap current and radio frequency current drive. Neutral beam injection and 3D non-resonant magnetic field can provide edge plasma rotation for stabilization of MHD and access to Quiescent H-mode. The estimated power exhaust is somewhat lower than that of ITER because of higher core radiation and stronger tilting of the divertor plates. FDF is capable of further developing all elements of AT physics, qualifying them for an advanced performance DEMO. The latest concept has accounted for realistic neutron shielding and divertor implementation. Self-consistent evolution of the transport profiles and equilibrium will quantify the stability and confinement required to meet the FNS mission.
NASA Astrophysics Data System (ADS)
Howard, N. T.; Holland, C.; White, A. E.; Greenwald, M.; Candy, J.
2016-01-01
The transport of heat in laboratory and astrophysical plasmas is dominated by the complex nonlinear dynamics of plasma turbulence. In magnetically confined plasmas used for fusion energy research, turbulence is responsible for cross-field transport that limits the performance of tokamak reactors. We report a set of novel gyrokinetic simulations that capture ion and electron-scale turbulence simultaneously, revealing the dynamics of cross-scale energy transfer and zonal flow modification that give rise to heat losses. Multi-scale simulations are required to match experimental ion and electron heat fluxes and electron profile stiffness, establishing the applicability of the newly discovered physics to experiment. Importantly, these results provide a likely explanation for the loss of electron heat from tokamak plasmas, the ‘great unsolved problem’ (Bachelor et al (2007 Plasma Sci. Technol. 9 312-87)) in plasma turbulence and the projected dominant loss channel in ITER.
Liu, Z. X.; Gao, X.; Liu, S. C.; Ding, S. Y.; Li, J. G.; Xia, T. Y.; Xu, X. Q.; Hughes, J. W.
2012-10-15
H-mode plasmas with ELM (edge localized mode) have been realized on experimental advanced superconducting tokamak (EAST) with 2.45 GHz low hybrid wave at P{sub LHW}{approx}1 MW in 2010. Data from EAST experiments including magnetic geometry, measured pressure profiles, and calculated current profiles are used to investigate the physics of ELM utilizing the BOUT++ code. Results from linear simulations show that the ELMs in EAST are dominated by resistive ballooning modes. When the Lundquist number (dimensionless ratio of the resistive diffusion time to the Alfven time) is equal to or less than 10{sup 7}, the resistive ballooning modes are found to become unstable in the ELMy H-mode plasma. For a fixed pedestal pressure profile, increasing plasma current generates more activities of low-n ELMs.
Rodrigues, Paulo; Bizarro, Joao P. S.
2007-09-21
For the first time, tokamak equilibria with negative toroidal current flowing in the plasma core are computed consistently with available measurements from typical current-hole discharges. The equilibrium reconstruction, which leads to non-nested configurations where a system of axisymmetric magnetic islands unfolds, yields an overall good agreement between the computed and experimental plasma-pressure profiles, together with an excellent fit to motional-Stark-effect data. Therefore, considering the accuracy limits of present-day experimental results, care must be exercised when ruling out the existence of tokamak equilibria with central toroidal-current reversal, particularly if relying on reconstruction tools that cannot cope with non-nested configurations.
The influence of the ion polarization current on magnetic island stability in a tokamak plasma
Fitzpatrick, R.; Waelbroeck, F. L.; Militello, F.
2006-12-15
The influence of the ion polarization current on the stability of a constant-{psi} magnetic island in a tokamak plasma is investigated numerically using a reduced two-fluid model in two-dimensional slab geometry. The polarization current is found to be negligibly small when the island is either too narrow or too wide. However, under certain circumstances, there exists an intermediate regime in which the polarization current is appreciable, and has a stabilizing influence on the island. This effect may account for the metastable nature of neoclassical tearing modes in tokamak plasmas.
Zonal flow modes in a tokamak plasma with dominantly poloidal mean flows
Zhou Deng
2010-10-15
The zonal flow eigenmodes in a tokamak plasma with dominantly poloidal mean flows are theoretically investigated. It is found that the frequencies of both the geodesic acoustic mode and the sound wave increase with respect to the poloidal Mach number. In contrast to the pure standing wave form in static plasmas, the density perturbations consist of a standing wave superimposed with a small amplitude traveling wave in the poloidally rotating plasma.
Kessel, C. E.; Poli, F. M.; Ghantous, K.; Gorelenkov, N. N.; Rensink, M. E.; Rognlien, T. D.; Snyder, P. B.; St. John, H.; Turnbull, A. D.
2015-01-01
Here, the advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, and triangularity of 0.63. The broadest pressure cases reached wall-stabilized β_{N} ~ 5.75, limited by n = 3 external kink mode requiring a conducting shell at b/a = 0.3, requiring plasma rotation, feedback, and/or kinetic stabilization. The medium pressure peaking case reaches β_{N} = 5.28 with B_{T} = 6.75, while the peaked pressure case reaches β_{N} < 5.15. Fast particle magnetohydrodynamic stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling shows that 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while >95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring ~1.1 MA of external current drive. This current is supplied with 5 MW of ion cyclotron radio frequency/fast wave and 40 MW of lower hybrid current drive. Electron cyclotron is most effective for safety factor control over ρ~0.2 to 0.6 with 20 MW. The pedestal density is ~0.9×10^{20}/m^{3}, and the temperature is ~4.4 keV. The H98 factor is 1.65, n/n_{Gr} = 1.0, and the ratio of net power to threshold power is 2.8 to 3.0 in the flattop.
Kessel, C. E.; Poli, F. M.; Ghantous, K.; Gorelenkov, N. N.; Rensink, M. E.; Rognlien, T. D.; Snyder, P. B.; St. John, H.; Turnbull, A. D.
2015-01-01
Here, the advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at an aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2, and triangularity of 0.63. The broadest pressure cases reached wall-stabilized βN ~ 5.75, limited by n = 3 external kink mode requiring a conducting shell at b/a = 0.3, requiring plasma rotation, feedback, and/or kinetic stabilization. The medium pressure peaking case reaches βN = 5.28 with BT = 6.75, while the peaked pressure case reaches βN < 5.15. Fast particle magnetohydrodynamic stability shows that themore » alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling shows that 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while >95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring ~1.1 MA of external current drive. This current is supplied with 5 MW of ion cyclotron radio frequency/fast wave and 40 MW of lower hybrid current drive. Electron cyclotron is most effective for safety factor control over ρ~0.2 to 0.6 with 20 MW. The pedestal density is ~0.9×1020/m3, and the temperature is ~4.4 keV. The H98 factor is 1.65, n/nGr = 1.0, and the ratio of net power to threshold power is 2.8 to 3.0 in the flattop.« less
Charles Kessel, et al
2014-03-05
The advanced physics and advanced technology tokamak power plant ARIES-ACT1 has a major radius of 6.25 m at aspect ratio of 4.0, toroidal field of 6.0 T, strong shaping with elongation of 2.2 and triangularity of 0.63. The broadest pressure cases reached wall stabilized βN ~ 5.75, limited by n=3 external kink mode requiring a conducting shell at b/a = 0.3, and requiring plasma rotation, feedback, and or kinetic stabilization. The medium pressure peaking case reached βN = 5.28 with BT = 6.75, while the peaked pressure case reaches βN < 5.15. Fast particle MHD stability shows that the alpha particles are unstable, but this leads to redistribution to larger minor radius rather than loss from the plasma. Edge and divertor plasma modeling show that about 75% of the power to the divertor can be radiated with an ITER-like divertor geometry, while over 95% can be radiated in a stable detached mode with an orthogonal target and wide slot geometry. The bootstrap current fraction is 91% with a q95 of 4.5, requiring about ~ 1.1 MA of external current drive. This current is supplied with 5 MW of ICRF/FW and 40 MW of LHCD. EC was examined and is most effective for safety factor control over ρ ~ 0.2-0.6 with 20 MW. The pedestal density is ~ 0.9x1020 /m3 and the temperature is ~ 4.4 keV. The H98 factor is 1.65, n/nGr = 1.0, and the net power to LH threshold power is 2.8- 3.0 in the flattop.
NASA Astrophysics Data System (ADS)
DeBono, Bryan Angelo
This thesis presents the systematic study of the multimode external kink mode structure and dynamics in the High-Beta Tokamak Extended-Pulse experiment (HBT-EP) when the plasma rotation is externally controlled using a source of toroidal momentum input. The capabilities of the HBT-EP tokamak to study rotation physics was greatly extended during a 2009--2010 major upgrade, when a new adjustable conducting wall, a high-power modular control coil array system, and an extensive set of 216 poloidal and radial magnetic sensors were installed on the machine. HBT-EP was additionally equipped with a biased edge electrode which made it possible to adjust the plasma ion and plasma magnetohydrodynamics (MHD) mode rotation frequencies by imparting an electromagnetic torque on the plasma. The design of this biased edge electrode, and its capability to torque the plasma is described. The rotation frequency of the helical kink modes was directly inferred from analysis of the magnetics dataset. To directly measure the plasma ion acceleration as the plasma was torqued by the biased electrode, a novel high-throughput and fast-response spectroscopic rotation diagnostic was installed on HBT-EP. This spectroscopic rotation diagnostic was designed to measure the velocity of He ions, therefore when conducting experiments using the spectroscopic rotation diagnostic a gas mixture of 90%D and 10%He was used. With its current power supplies the bias probe is capable of accelerating the primary m/n=3/1 helical kink mode (which has a natural rotation frequency between +7→+9kHz) to somewhere between -50kHz→+25kHz depending on the probe bias. At a probe voltage of +175V the He impurity ions were seen to accelerate by 3km/sec. Biorthogonal decomposition (BD) analysis was applied to the large magnetics dataset and used to determine the multimode m/n spectrum of the helical kink modes present in HBT-EP. The dominant helicities present as revealed by the BD are the m/n=3/1 and m/n=6/2 modes
Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak
NASA Astrophysics Data System (ADS)
Li, Y. L.; Xu, G. S.; Tritz, K.; Zhu, Y. B.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L.
2015-12-01
A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.
Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak.
Li, Y L; Xu, G S; Tritz, K; Zhu, Y B; Wan, B N; Lan, H; Liu, Y L; Wei, J; Zhang, W; Hu, G H; Wang, H Q; Duan, Y M; Zhao, J L; Wang, L; Liu, S C; Ye, Y; Li, J; Lin, X; Li, X L
2015-12-01
A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks. PMID:26724032
Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak
Li, Y. L.; Xu, G. S.; Wan, B. N.; Lan, H.; Liu, Y. L.; Wei, J.; Zhang, W.; Hu, G. H.; Wang, H. Q.; Duan, Y. M.; Zhao, J. L.; Wang, L.; Liu, S. C.; Ye, Y.; Li, J.; Lin, X.; Li, X. L.; Tritz, K.; Zhu, Y. B.
2015-12-15
A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structures were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.
Plasma Processing of Advanced Materials
Heberlein, Joachim, V.R.; Pfender, Emil; Kortshagen, Uwe
2005-02-28
Plasma Processing of Advanced Materials The project had the overall objective of improving our understanding of the influences of process parameters on the properties of advanced superhard materials. The focus was on high rate deposition processes using thermal plasmas and atmospheric pressure glow discharges, and the emphasis on superhard materials was chosen because of the potential impact of such materials on industrial energy use and on the environment. In addition, the development of suitable diagnostic techniques was pursued. The project was divided into four tasks: (1) Deposition of superhard boron containing films using a supersonic plasma jet reactor (SPJR), and the characterization of the deposition process. (2) Deposition of superhard nanocomposite films in the silicon-nitrogen-carbon system using the triple torch plasma reactor (TTPR), and the characterization of the deposition process. (3) Deposition of films consisting of carbon nanotubes using an atmospheric pressure glow discharge reactor. (4) Adapting the Thomson scattering method for characterization of atmospheric pressure non-uniform plasmas with steep spatial gradients and temporal fluctuations. This report summarizes the results.
Physics Basis for the Advanced Tokamak Fusion Power Plant ARIES-AT
S.C. Jardin; C.E. Kessel; T.K. Mau; R.L. Miller; F. Najmabadi; V.S. Chan; M.S. Chu; R. LaHaye; L.L. Lao; T.W. Petrie; P. Politzer; H.E. St. John; P. Snyder; G.M. Staebler; A.D. Turnbull; W.P. West
2003-10-07
The advanced tokamak is considered as the basis for a fusion power plant. The ARIES-AT design has an aspect ratio of A always equal to R/a = 4.0, an elongation and triangularity of kappa = 2.20, delta = 0.90 (evaluated at the separatrix surface), a toroidal beta of beta = 9.1% (normalized to the vacuum toroidal field at the plasma center), which corresponds to a normalized beta of bN * 100 x b/(I(sub)P(MA)/a(m)B(T)) = 5.4. These beta values are chosen to be 10% below the ideal-MHD stability limit. The bootstrap-current fraction is fBS * I(sub)BS/I(sub)P = 0.91. This leads to a design with total plasma current I(sub)P = 12.8 MA, and toroidal field of 11.1 T (at the coil edge) and 5.8 T (at the plasma center). The major and minor radii are 5.2 and 1.3 m, respectively. The effects of H-mode edge gradients and the stability of this configuration to non-ideal modes is analyzed. The current-drive system consists of ICRF/FW for on-axis current drive and a lower-hybrid system for off-axis. Tran sport projections are presented using the drift-wave based GLF23 model. The approach to power and particle exhaust using both plasma core and scrape-off-layer radiation is presented.
Flow shear induced fluctuation suppression in finite aspect ratio shaped tokamak plasma
Hahm, T.S.; Burrell, K.H.
1995-01-01
The suppression of turbulence by the E {times} B flow shear and parallel flow shear is studied in an arbitrary shape finite aspect ratio tokamak plasma using the two point nonlinear analysis previously utilized in a high aspect rat& tokamak plasma. The result shows that only the E {times} B flow shear is responsible for the suppression of flute-like fluctuations. This suppression occurs regardless of the plasma rotation direction and is therefore, relevant for the VH mode plasma core as well as for the H mode plasma edge. Experimentally observed in-out asymmetry of fluctuation reduction behavior can be addressed in the context of flux expansion and magnetic field pitch variation on a given flux surface. The adverse effect of neutral particles on confinement improvement is also discussed in the context of the charge exchange induced parallel momentum damping.
Generation of two-column helicon plasma on KAIST-TOKAMAK
NASA Astrophysics Data System (ADS)
Jeon, S. J.; Huh, S. W.; Kim, J.; Lee, T. S.; Moon, S. Y.; Choe, W.; Choi, D. I.
2000-10-01
Industrial plasma application studies reveal that helicon waves provide high ionization rate even at modest rf input power. This suggests that helicon waves be effectively used for plasma pre-ionization/startup, and plasma heating in a tokamak. The two-column helicon plasma was produced with a Nagoya type ¥2 antenna which was modified for toroidal geometry of KAIST-TOKAMAK. The observed two columns locate at the same major radius and they move outward as toroidal magnetic field increases. In addition to the 2D image captured by a CCD camera, an 8-channel Langmuir probe array is used to measure the density profile. Parallel wave number is measured by magnetic pickup probes and a phase detector in order to study wave generation and propagation inside the plasma.
A flexible software design to determine the plasma boundary in Damavand tokamak
NASA Astrophysics Data System (ADS)
Ghadiri, Rasoul; Sadeghi, Yahya; Esteki, Mohammad Hossein
2014-06-01
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.
Thermal loads on tokamak plasma facing components during normal operation and disruptions
NASA Astrophysics Data System (ADS)
McGrath, Robert T.
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 component and of the magnetic flux surface is very important in the modeling. Experimental measurements show that misalignment of component armor 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 armor and structure due to localized impact of runaway electrons.
Rotation of weakly collisional plasmas in tokamaks, operated with Alfv{acute e}n waves
Tsypin, V.S.; Elfimov, A.G.; de Azevedo, C.A.; de Assis, A.S.
1996-12-01
The effect of the kinetic Alfv{acute e}n waves on weakly collisional plasma rotation in tokamaks has been studied for the plateau and banana regimes. The quasistationary rotation velocities and radial electric field have been found. The estimation of these quantities for the Phaedrus-T tokamak [S. Wukitch {ital et} {ital al}., Phys. Rev. Lett. {bold 77}, 294 (1996)] and for the Joint European Torus (JET) [A. Fasoli {ital et} {ital al}., Nucl. Fusion, {bold 36}, 258 (1996)] has been presented. It is shown that the kinetic Alfv{acute e}n waves, which are needed for current drive, change weakly the quasistationary rotation velocities and radial electric field, as found from the experimental data of these tokamaks. In conditions with increased rf power, the plasma rotation and radial electric field can essentially grow up. {copyright} {ital 1996 American Institute of Physics.}
Effect of a poloidal electric field on neoclassical transport in a multispecies tokamak plasma
Indireshkumar, K.; Stacey, W.M. Jr.
1992-12-01
The effects of a poloidal potential variation of or der c{var_epsilon}, heating or neutral beam injection, upon neoclassical particle transport and plasma current are studied theoretically, for a realistic tokamak plasma with significant impurity content. Using an approximate collision operator, an analytic procedure is employed to calculate the transport coefficients in the low collisionality regime for a large aspect ratio tokamak. In the presence of carbon impurity, the ion diffusion coefficients are generally found to increase by a factor of {approximately} 2. Inclusion of the effects of a poloidal electric field is found to result in an increase in the bootstrap current if the potential on the outside of the tokamak is greater than that on the inside (as during ICRH or NBI) and the density profiles are more peaked than roughly the square root of the temperature profiles.
Effect of a poloidal electric field on neoclassical transport in a multispecies tokamak plasma
Indireshkumar, K.; Stacey, W.M. Jr.
1992-12-01
The effects of a poloidal potential variation of or der c[var epsilon], heating or neutral beam injection, upon neoclassical particle transport and plasma current are studied theoretically, for a realistic tokamak plasma with significant impurity content. Using an approximate collision operator, an analytic procedure is employed to calculate the transport coefficients in the low collisionality regime for a large aspect ratio tokamak. In the presence of carbon impurity, the ion diffusion coefficients are generally found to increase by a factor of [approximately] 2. Inclusion of the effects of a poloidal electric field is found to result in an increase in the bootstrap current if the potential on the outside of the tokamak is greater than that on the inside (as during ICRH or NBI) and the density profiles are more peaked than roughly the square root of the temperature profiles.
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.
2015-10-30
New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong E x B shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offering one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. In conclusion, this predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
Wang, W. X.; Ethier, S.; Ren, Y.; Kaye, S.; Chen, J.; Startsev, E.; Lu, Z.
2015-10-30
New turbulence contributions to plasma transport and confinement in the spherical tokamak (ST) regime are identified through nonlinear gyrokinetic simulations. The drift wave Kelvin-Helmholtz (KH) mode characterized by intrinsic mode asymmetry is shown to drive significant ion thermal transport in strongly rotating national spherical torus experiment (NSTX) L-modes. The long wavelength, quasi-coherent dissipative trapped electron mode (TEM) is destabilized in NSTX H-modes despite the presence of strong E x B shear, providing a robust turbulence source dominant over collisionless TEM. Dissipative trapped electron mode (DTEM)-driven transport in the NSTX parametric regime is shown to increase with electron collision frequency, offeringmore » one possible source for the confinement scaling observed in experiments. There exists a turbulence-free regime in the collision-induced collisionless trapped electron mode to DTEM transition for ST plasmas. In conclusion, this predicts a natural access to a minimum transport state in the low collisionality regime that future advanced STs may cover.« less
Equilibrium Plasma Position Control for a Large Tokamak Using Modern Control Theory
NASA Astrophysics Data System (ADS)
Fukunishi, Kohyu; Saito, Seiji; Ogata, Atsushi; Ninomiya, Hiromasa
1980-09-01
Optimal control techniques are applied to maintain the plasma in its equilibrium position in a large tokamak. The application of the state space equation to plasma position control is also discussed. Optimal controls with states, which are plasma current, OH coil current and vertical field current, and integrated plasma displacement feedbacks are formulated as linear, time invariant expressions with quadratic performance indices. Effective plasma position control was obtained with integral state feedback in computer simulations for the JT-60. These control techniques will be applied to the JT-60.
NASA Astrophysics Data System (ADS)
von Nessi, G. T.; Hole, M. J.; The MAST Team
2014-11-01
We present recent results and technical breakthroughs for the Bayesian inference of tokamak equilibria using force-balance as a prior constraint. Issues surrounding model parameter representation and posterior analysis are discussed and addressed. These points motivate the recent advancements embodied in the Bayesian Equilibrium Analysis and Simulation Tool (BEAST) software being presently utilized to study equilibria on the Mega-Ampere Spherical Tokamak (MAST) experiment in the UK (von Nessi et al 2012 J. Phys. A 46 185501). State-of-the-art results of using BEAST to study MAST equilibria are reviewed, with recent code advancements being systematically presented though out the manuscript.
TPX diagnostics for tokamak operation, plasma control and machine protection
Edmonds, P.H.; Medley, S.S.; Young, K.M.
1995-08-01
The diagnostics for TPX are at an early design phase, with emphasis on the diagnostic access interface with the major tokamak components. Account has to be taken of the very severe environment for diagnostic components located inside the vacuum vessel. The placement of subcontracts for the design and fabrication of the diagnostic systems is in process.
Cherenkov-type diagnostics of fast electrons within tokamak plasmas
NASA Astrophysics Data System (ADS)
Jakubowski, Lech; Sadowski, Marek J.; Zebrowski, Jaroslaw; Malinowski, Karol; Rabinski, Marek; Jakubowski, Marcin J.; Mirowski, Robert
2014-05-01
This paper presents a summary of the most important results of fast electron measurements performed so far within different tokamaks by means of Cherenkov-type detectors. In the ISTTOK tokamak (IPFN, IST, Lisboa, Portugal), two measuring heads were applied, each equipped with four radiators made of different types of alumina-nitrate poly-crystals. A two-channel measuring head equipped with diamond radiators was also used. Within the COMPASS tokamak (IPP AS CR, Prague, Czech Republic) some preliminary measurements have recently been performed by means of a new single-channel Cherenkov-type detector. The experimental data from the TORE SUPRA tokamak (CEA, IFRM, Cadarache, France), which were collected by means of a DENEPR-2 probe during two recent experimental campaigns, have been briefly analyzed. A new Cherenkov probe (the so-called DENEPR-3) has been mounted within the TORE SUPRA machine, but the electron measurements could not be performed because of the failure of this facility. Some conclusions concerning the fast electron emission are presented.
2-D Imaging of Electron Temperature in Tokamak Plasmas
T. Munsat; E. Mazzucato; H. Park; C.W. Domier; M. Johnson; N.C. Luhmann Jr.; J. Wang; Z. Xia; I.G.J. Classen; A.J.H. Donne; M.J. van de Pol
2004-07-08
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.
Advances in Dust Detection and Removal for Tokamaks
NASA Astrophysics Data System (ADS)
Campos, A.; Skinner, C. H.; Roquemore, A. L.; Leisure, J. O. V.; Wagner, S.
2008-11-01
Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. An electrostatic dust detector[1] developed in the laboratory is being applied to NSTX. In the tokamak environment, large particles or fibres can fall on the grid potentially causing a permanent short. We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments with varying nozzle designs, backing pressures, puff durations, and exit flow orientations have obtained an optimal configuration that effectively removes particles from a 25 cm^2 area. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tripolar grid of fine interdigitated traces has been designed that generates an electrostatic travelling wave for conveying dust particles to a ``drain.'' First trials have shown particle motion in optical microscope images. [1] C. H. Skinner et al., J. Nucl. Mater., 376 (2008) 29.
Overview of JT-60U results towards the establishment of advanced tokamak operation
NASA Astrophysics Data System (ADS)
Oyama, N.; JT-60 Team
2009-10-01
Recent JT-60U experimental results towards the establishment of advanced tokamak (AT) operation are reviewed. We focused on the further expansion of the operational regime of AT plasmas towards higher βN regime with wall stabilization. After the installation of ferritic steel tiles in 2005, the high power heating in a large plasma cross-section in which the wall stabilization is expected has been possible. In 2007, the modification of power supply of NBIs improved the flexibility of the heating profile in long-pulse plasmas. The investigation of key physics issues for the establishment of steady-state AT operation is also in progress using new diagnostics and improved heating systems. In weak magnetic shear plasma, high βN ~ 3 exceeding the ideal MHD limit without a conducting wall ( \\beta_N^{{\\scriptsize{\\mbox{no-wall}}}} ) is sustained for ~5 s (~3τR) with RWM stabilization by a toroidal rotation at the q = 2 surface. External current drivers of negative-ion based NB and lower-hybrid waves together with a large bootstrap current fraction (fBS) of 0.5 can sustain the whole plasma current of 0.8 MA for 2 s (1.5τR). In reversed magnetic shear plasma, high βN ~ 2.7 (βp ~ 2.3) exceeding \\beta_N^{{\\scriptsize{\\mbox{no-wall}}}} with qmin ~ 2.4 (q95 ~ 5.3), HH98(y,2) ~ 1.7 and fBS ~ 0.9 is obtained with wall stabilization. These plasma parameters almost satisfy the requirement of ITER steady-state scenario. In long-pulse plasmas with positive magnetic shear, a high βNHH98(y,2) of 2.6 with βN ~ 2.6 and HH98(y,2) ~ 1 is sustained for 25 s, significantly longer than the current diffusion time (~14τR) without neoclassical tearing modes (NTMs). A high G-factor, \\beta_NH_{89P}/q_{95}^{2} (a major of fusion gain), of 0.54 and a large fBS > 0.43 are suitable for ITER hybrid operation scenario. Based on the plasma for ITER hybrid operation scenario, the high βN of 2.1 with good thermal plasma confinement of HH98(y,2) > 0.85 is sustained for longer than 12 s at
Limiter/vacuum system for plasma impurity control and exhaust in tokamaks
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.
Effects of orbit squeezing on poloidal mass flow and bootstrap current in tokamak plasmas
Shaing, K.C. ); Hsu, C.T. ); Hazeltine, R.D. )
1994-10-01
It is shown, by solving the drift kinetic equation, that the asymptotic values of the poloidal mass flow and the bootstrap current in the banana regime of large-aspect-ratio tokamak plasmas are not affected by orbit squeezing. However, because the definition of ion collisionality [upsilon][sub *[ital i
Resonance parallel viscosity in the banana regime in poloidally rotating tokamak plasmas
Shaing, K.C.; Hsu, C.T.; Dominguez, N. )
1994-05-01
Parallel viscosity in the banana regime in a poloidally ([bold E][times][bold B]) rotating tokamak plasma is calculated to include the effects of orbit squeezing and to allow the poloidal [bold E][times][bold B] Mach number [ital M][sub [ital p
Role of neutral gas in scrape-off layer tokamak plasma
Bisai, N.; Jha, R.; Kaw, P. K.
2015-02-15
Neutral gas in scrape-off layer of tokamak plasma plays an important role as it can modify the plasma turbulence. In order to investigate this, we have derived a simple two-dimensional (2D) model that consists of electron continuity, quasi-neutrality, and neutral gas continuity equations using neutral gas ionization and charge exchange processes. Simple 1D profile analysis predicts neutral penetration depth into the plasma. Growth rate obtained from the linear theory has been presented. The 2D model equations have been solved numerically. It is found that the neutral gas reduces plasma fluctuations and shifts spectrum of the turbulence towards lower frequency side. The neutral gas fluctuation levels have been presented. The numerical results have been compared with Aditya tokamak experiments.
Influence of collisions on parametric instabilities induced by lower hybrid waves in tokamak plasmas
NASA Astrophysics Data System (ADS)
Castaldo, C.; Di Siena, A.; Fedele, R.; Napoli, F.; Amicucci, L.; Cesario, R.; Schettini, G.
2016-01-01
Parametric instabilities induced at the plasma edge by lower hybrid wave power externally coupled to tokamak plasmas have, via broadening of the antenna spectrum, strong influence on the power deposition and current drive in the core. For modeling the parametric instabilities at the tokamak plasma edge in lower hybrid current drive experiments, the effect of the collisions has been neglected so far. In the present work, a specific collisional parametric dispersion relation, useful to analyze these nonlinear phenomena near the lower hybrid antenna mouth, is derived for the first time, based on a kinetic model. Numerical solutions show that in such cold plasma regions the collisions prevent the onset of the parametric instabilities. This result is important for present lower hybrid current drive experiments, as well as in fusion reactor scenarios.
The simulation of L-H transition in tokamak plasma using MMM95 transport model
NASA Astrophysics Data System (ADS)
Intharat, P.; Chatthong, B.; Onjun, T.; Poolyarat, N.; Picha, R.
2015-05-01
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.
High βp plasma formation using off-axis ECCD in Ohmic heated plasma in the spherical tokamak QUEST
NASA Astrophysics Data System (ADS)
Mishra, Kishore; Zushi, H.; Idei, H.; Hasegawa, M.; Hanada, K.
2015-03-01
High poloidal beta (ɛβp ~ 1) operation in steady state condition in tokamaks is of great interest and has previously been demonstrated using NBI, LHCD and low current (Ip) plasma for a short time (<0.5 s). A very few experiments however, have been performed towards the investigation of highest obtainable βp in tokamak plasma. In this work we report the first result of high βp production and its sustainment though an off axis ECCD at two different frequencies (fundamental and second harmonic) in Ohmic (OH) target plasma. With application of ECCD, plasma βp increased to encounter an equilibrium limit and the standard limiter configuration is transformed to an Inboard Poloidal field Null (IPN) configuration. Both off-axis and on-axis ECCD is studied and found to have some distinctive features, which are discussed in this paper.
Energy Confinement of High-Density Pellet-Fueled Plasmas in the Alcator C Tokamak
NASA Astrophysics Data System (ADS)
Greenwald, M.; Gwinn, D.; Milora, S.; Parker, J.; Parker, R.; Wolfe, S.; Besen, M.; Camacho, F.; Fairfax, S.; Fiore, C.; Foord, M.; Gandy, R.; Gomez, C.; Granetz, R.; Labombard, B.; Lipschultz, B.; Lloyd, B.; Marmar, E.; McCool, S.; Pappas, D.; Petrasso, R.; Pribyl, P.; Rice, J.; Schuresko, D.; Takase, Y.; Terry, J.; Watterson, R.
1984-07-01
A series of pellet-fueling experiments has been carried out on the Alcator C tokamak. High-speed hydrogen pellets penetrate to within a few centimeters of the magnetic axis, raise the plasma density, and produce peaked density profiles. Energy confinement is observed to increase over similar discharges fueled only by gas puffing. In this manner record values of electron density, plasma pressure, and Lawson number (n τ) have been achieved.
NASA Astrophysics Data System (ADS)
Minashin, P. V.; Kukushkin, A. B.
2015-03-01
A method of spectroscopic diagnostics of the average perpendicular-to-magnetic-field momentum of the superthermal component of the electron velocity distribution (EVD), based on the high-number-harmonic electron cyclotron (EC) radiation, is suggested for nuclear fusion-reactor plasmas under condition of a strong auxiliary heating (e.g. in tokamak DEMO, a next step after tokamak ITER). The method is based on solving an inverse problem for reconstruction of the EVD in parallel and perpendicular-to-magnetic-field components of electron momentum at high and moderate energies responsible for the emission of the high-number-harmonic EC radiation.
Development of magnetohydrodynamic modes during sawteeth in tokamak plasmas
Firpo, M.-C.; Ettoumi, W.; Farengo, R.; Ferrari, H. E.; García-Martínez, P. L.; Lifschitz, A. F.
2013-07-15
A dynamical analysis applied to a reduced resistive magnetohydrodynamics model is shown to explain the chronology of the nonlinear destabilization of modes observed in tokamak sawteeth. A special emphasis is put on the nonlinear self-consistent perturbation of the axisymmetric m = n = 0 mode that manifests through the q-profile evolution. For the very low fusion-relevant resistivity values, the q-profile is shown to remain almost unchanged on the early nonlinear timescale within the central tokamak region, which supports a partial reconnection scenario. Within the resistive region, indications for a local flattening or even a local reversed-shear of the q-profile are given. The impact of this ingredient in the occurrence of the sawtooth crash is discussed.
Electron Transport by Radio Frequency Waves in Tokamak Plasmas
Ram, A. K.; Kominis, Y.; Hizanidis, K.
2009-11-26
A relativistic kinetic description for momentum and spatial diffusion of electrons by radio frequency (RF) waves and non-axisymmetric magnetic field perturbations in a tokamak is formulated. The Lie perturbation technique is used to obtain a non-singular, time dependent evolution equation for resonant and non-resonant electron diffusion in momentum space and diffusion in configuration space. The kinetic equation for the electron distribution function is different from the usual quasilinear equations as it includes interactions that are non-Markovian. It is suitable for studying wave-particle interaction in present tokamaks and in ITER. A primary goal of RF waves, and, in particular, of electron cyclotron waves, in ITER is to control instabilities like the neoclassical tearing mode (NTM). Non-axisymmetric effects due to NTMs are included in the kinetic formalism.
Development of magnetohydrodynamic modes during sawteeth in tokamak plasmas
NASA Astrophysics Data System (ADS)
Firpo, M.-C.; Ettoumi, W.; Farengo, R.; Ferrari, H. E.; García-Martínez, P. L.; Lifschitz, A. F.
2013-07-01
A dynamical analysis applied to a reduced resistive magnetohydrodynamics model is shown to explain the chronology of the nonlinear destabilization of modes observed in tokamak sawteeth. A special emphasis is put on the nonlinear self-consistent perturbation of the axisymmetric m = n = 0 mode that manifests through the q-profile evolution. For the very low fusion-relevant resistivity values, the q-profile is shown to remain almost unchanged on the early nonlinear timescale within the central tokamak region, which supports a partial reconnection scenario. Within the resistive region, indications for a local flattening or even a local reversed-shear of the q-profile are given. The impact of this ingredient in the occurrence of the sawtooth crash is discussed.
Kinetic effects in the conversion of fast waves in pre-heated, low aspect ratio tokamak plasmas
NASA Astrophysics Data System (ADS)
Kommoshvili, K.; Cuperman, S.; Bruma, C.
2003-03-01
Kinetic effects in the conversion of fast waves to Alfvèn waves and their subsequent deposition in low aspect ratio (spherical) tokamaks (LARTs) have been investigated theoretically. More specifically, we have considered the consequences of incorporation of kinetic effects in the electron parallel (to the ambient magnetic field) dynamics derived by following the drift-tearing mode analysis of Chen et al (Chen L, Rutherford P H and Tang W M 1977 Phys. Rev. Lett. 39 460), and particle-conserving Krook collision operator for the passing electrons involved (Mett R R and Mahajan S M 1992 Phys. Fluids B 4 2885). The perpendicular plasma dynamics is described by a quite general resistive two-fluid (2F) model based dielectric tensor-operator (Cuperman S, Bruma C and Komoshvili K 2002 Solution of the resistive 2F wave equations for Alfvènic modes in spherical tokamak plasmas J. Plasma Phys. accepted for publication). The full-wave electromagnetic equations, formulated in terms of the vector and scalar potentials, have been solved by the aid of an advanced finite elements numerical code (Sewell G 1993 Adv. Eng. Software 17 105). Detailed solutions of the full-wave equations are obtained and compared with those corresponding to a pure resistive 2F model, this, for the illustrative pre-heated START-type device (Sykes 1994). Our results quantitatively confirm the general theory of the conversion of fast waves with subsequent power dissipation for the conditions of spherical tokamaks thus providing the required auxilliary energy source for the succesful operation of LARTs. Moreover, these results indicate the absolute necessity of using a full model for the parallel electron dynamics, i.e. including both kinetic and collisional effects.
Numerical study of Alfvén eigenmodes in the Experimental Advanced Superconducting Tokamak
Hu, Youjun; Li, Guoqiang; Yang, Wenjun; Zhou, Deng; Ren, Qilong; Gorelenkov, N. N.; Cai, Huishan
2014-05-15
Alfvén eigenmodes in up-down asymmetric tokamak equilibria are studied by a new magnetohydrodynamic eigenvalue code. The code is verified with the NOVA code for the Solovév equilibrium and then is used to study Alfvén eigenmodes in a up-down asymmetric equilibrium of the Experimental Advanced Superconducting Tokamak. The frequency and mode structure of toroidicity-induced Alfvén eigenmodes are calculated. It is demonstrated numerically that up-down asymmetry induces phase variation in the eigenfunction across the major radius on the midplane.
High kinetic energy plasma jet generation and its injection into the Globus-M spherical tokamak
NASA Astrophysics Data System (ADS)
Voronin, A. V.; Gusev, V. K.; Petrov, Yu. V.; Sakharov, N. V.; Abramova, K. B.; Sklyarova, E. M.; Tolstyakov, S. Yu.
2005-09-01
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 1022 m-3, total numbers of accelerated particles (1-5) × 1019, 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 (<0.5 ms) at all spatial points up to a radius rap 0.3a. The plasma particle inventory increase by ~50% (from 0.65 × 1019 to 1 × 1019) did not result in plasma degradation.
Particle Control and Plasma Performance in the Lithium Tokamak Experiment (LTX)
Richard Majeski, et. al.
2013-02-21
The Lithium Tokamak eXperiment (LTX) is a small, low aspect ratio tokamak, which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350{degree}C. Several gas fueling systems, including supersonic gas injection, and molecular cluster injection have been studied, and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 msec. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 msec. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak - thin, evaporated, liquefied coatings of lithium - does not produce an adequately clean surface.
NASA Astrophysics Data System (ADS)
Moreau, D.; Artaud, J. F.; Ferron, J. R.; Holcomb, C. T.; Humphreys, D. A.; Liu, F.; Luce, T. C.; Park, J. M.; Prater, R.; Turco, F.; Walker, M. L.
2015-06-01
This paper shows that semi-empirical data-driven models based on a two-time-scale approximation for the magnetic and kinetic control of advanced tokamak (AT) scenarios can be advantageously identified from simulated rather than real data, and used for control design. The method is applied to the combined control of the safety factor profile, q(x), and normalized pressure parameter, βN, using DIII-D parameters and actuators (on-axis co-current neutral beam injection (NBI) power, off-axis co-current NBI power, electron cyclotron current drive power, and ohmic coil). The approximate plasma response model was identified from simulated open-loop data obtained using a rapidly converging plasma transport code, METIS, which includes an MHD equilibrium and current diffusion solver, and combines plasma transport nonlinearity with 0D scaling laws and 1.5D ordinary differential equations. The paper discusses the results of closed-loop METIS simulations, using the near-optimal ARTAEMIS control algorithm (Moreau D et al 2013 Nucl. Fusion 53 063020) for steady state AT operation. With feedforward plus feedback control, the steady state target q-profile and βN are satisfactorily tracked with a time scale of about 10 s, despite large disturbances applied to the feedforward powers and plasma parameters. The robustness of the control algorithm with respect to disturbances of the H&CD actuators and of plasma parameters such as the H-factor, plasma density and effective charge, is also shown.
ADVANCES IN DUST DETECTION AND REMOVAL FOR TOKAMAKS
Campos, A.; Skinner, C.H.
2009-01-01
Dust diagnostics and removal techniques are vital for the safe operation of next step fusion devices such as ITER. In the tokamak environment, large particles or fi bers can fall on the electrostatic detector potentially causing a permanent short. An electrostatic dust detector developed in the laboratory is being applied to the National Spherical Torus Experiment (NSTX). We report on the development of a gas puff system that uses helium to clear such particles from the detector. Experiments at atmospheric pressure with varying nozzle designs, backing pressures, puff durations and exit fl ow orientations have given an optimal confi guration that effectively removes particles from a 25 cm² area. Similar removal effi ciencies were observed under a vacuum base pressure of 1 mTorr. Dust removal from next step tokamaks will be required to meet regulatory dust limits. A tri-polar grid of fi ne interdigitated traces has been designed that generates an electrostatic traveling wave for conveying dust particles to a “drain.” First trials with only two working electrodes have shown particle motion in optical microscope images.
Transport and confinement in the Mega Ampère Spherical Tokamak (MAST) plasma
NASA Astrophysics Data System (ADS)
Akers, R. J.; Ahn, J. W.; Antar, G. Y.; Appel, L. C.; Applegate, D.; Brickley, C.; Bunting, C.; Carolan, P. G.; Challis, C. D.; Conway, N. J.; Counsell, G. F.; Dendy, R. O.; Dudson, B.; Field, A. R.; Kirk, A.; Lloyd, B.; Meyer, H. F.; Morris, A. W.; Patel, A.; Roach, C. M.; Rohzansky, V.; Sykes, A.; Taylor, D.; Tournianski, M. R.; Valovi, M.; Wilson, H. R.; Axon, K. B.; Buttery, R. J.; Ciric, D.; Cunningham, G.; Dowling, J.; Dunstan, M. R.; Gee, S. J.; Gryaznevich, M. P.; Helander, P.; Keeling, D. L.; Knight, P. J.; Lott, F.; Loughlin, M. J.; Manhood, S. J.; Martin, R.; McArdle, G. J.; Price, M. N.; Stammers, K.; Storrs, J.; Walsh, M. J.; MAST, the; NBI Team
2003-12-01
A combination of recently installed state-of-the-art imaging and profile diagnostics, together with established plasma simulation codes, are providing for the first time on Mega Ampère Spherical Tokamak (MAST) the tools required for studying confinement and transport, from the core through to the plasma edge and scrape-off-layer (SOL). The H-mode edge transport barrier is now routinely turned on and off using a combination of poloidally localized fuelling and fine balancing of the X-points. Theory, supported by experiment, indicates that the edge radial electric field and toroidal flow velocity (thought to play an important role in H-mode access) are largest if gas fuelling is concentrated at the inboard side. H-mode plasmas show predominantly type III ELM characteristics, with confinement HH factor (w.r.t. scaling law IPB98[y, 2]) around ~1.0. Combining MAST H-mode data with the International Tokamak Physics Activities (ITPA) analyses, results in an L H power threshold scaling proportional to plasma surface area (rather than PLH ~ R2). In addition, MAST favours an inverse aspect ratio scaling PLH ~ egr0.5. Similarly, the introduction of type III ELMing H-mode data to the pedestal energy regression analysis introduces a scaling Wped ~ egr-2.13 and modifies the exponents on R, BT and kgr. Preliminary TRANSP simulations indicate that ion and electron thermal diffusivities in ELMing H-mode approach the ion-neoclassical level in the half-radius region of the plasma with momentum diffusivity a few times lower. Linear flux-tube ITG and ETG microstability calculations using GS2 offer explanations for the near-neoclassical ion diffusivity and significantly anomalous electron diffusivity seen on MAST. To complement the baseline quasi-steady-state H-mode, newly developed advanced regimes are being explored. In particular, 'broad' internal transport barriers (ITBs) have been formed using techniques developed at conventional aspect ratio. Electron and ion energy diffusivities
Spectroscopy and atomic physics of highly ionized Cr, Fe, and Ni for tokamak plasmas
NASA Technical Reports Server (NTRS)
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.
Can a Penning ionization discharge simulate the tokamak scrape-off plasma conditions?
NASA Technical Reports Server (NTRS)
Finkenthal, M.; Littman, A.; Stutman, D.; Kovnovich, S.; Mandelbaum, P.; Schwob, J. L.; Bhatia, A. K.
1990-01-01
The tokamak scrape-off (the region between the vacuum vessel wall and the magnetically confined fusion plasma edge), represents a source/sink for the hot fusion plasma. The electron densities and temperatures are in the ranges 10 to the 11th - 10 to the 13th/cu cm and 1-40 eV, respectively (depending on the size, magnetic field intensity and configuration, plasma current, etc). In the work reported, the electron temperature and density have been estimated in a Penning ionization discharge by comparing its spectroscopic emission in the VUV with that predicted by a collisional radiative model. An attempt to directly compare this emission with that of the tokamak edge is briefly described.
Peeling-off of the external kink modes at tokamak plasma edge
Zheng, L. J.; Furukawa, M.
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.
Resistive wall mode and neoclassical tearing mode coupling in rotating tokamak plasmas
NASA Astrophysics Data System (ADS)
McAdams, Rachel; Wilson, H. R.; Chapman, I. T.
2013-08-01
A model system of equations has been derived to describe a toroidally rotating tokamak plasma, unstable to resistive wall modes (RWMs) and metastable to neoclassical tearing modes (NTMs), using a linear RWM model and a nonlinear NTM model. If no wall is present, the NTM growth shows the typical threshold/saturation island widths, whereas a linearly unstable kink mode grows exponentially in this model plasma system. When a resistive wall is present, the growth of the linearly unstable RWM is accelerated by an unstable island: a form of coupled RWM-NTM mode. Crucially, this coupled system has no threshold island width, giving the impression of a triggerless NTM, observed in high beta tokamak discharges. Increasing plasma rotation at the island location can mitigate its growth, decoupling the modes to yield a conventional RWM with no threshold width.
Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M
2012-01-01
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. PMID:22233634
Can tokamaks PFC survive a single event of any plasma instabilities?
NASA Astrophysics Data System (ADS)
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.
Analysis of measurement errors for Thomson diagnostics of non-Maxwellian plasmas in tokamak reactors
NASA Astrophysics Data System (ADS)
Sdvizhenskii, P. A.; Kukushkin, A. B.; Kurskiev, G. S.; Mukhin, E. E.; Bassan, M.
2016-01-01
The study is stimulated by the expected noticeable deviation of the electron velocity distribution function (eVDF) from a Maxwellian under condition of a strong auxiliary heating of electron plasmas in tokamak-reactors. The key principles of accuracy estimation of the Thomson scattering diagnostic of non-Maxwellian plasmas in tokamak-reactors are presented. The algorithm extends the conventional approach to the assessment of non-Maxwellian plasmas measurements errors for a broad class of deviations of the eVDF from a Maxwellian. The algorithm is based on solving the inverse problem many times to determine main parameters of the eVDF with allowance for all possible sources of error and statistical variation of the input parameters of the problem. The method is applied to a preliminary analysis of the advantages of the formerly suggested use of various wavelengths of probing laser radiation in the Thomson diagnostics of non-Maxwellian plasma on the example of the core plasma Thomson scattering diagnostic system which is under design for ITER tokamak. The results obtained confirm the relevance of the diversification of the probing laser radiation wavelength.
Coupled two-dimensional edge plasma and neutral gas modeling of tokamak scrape-off-layers
Maingi, R.
1992-08-01
The objective of this study is to devise a detailed description of the tokamak scrape-off-layer (SOL), which includes the best available models of both the plasma and neutral species and the strong coupling between the two in many SOL regimes. A good estimate of both particle flux and heat flux profiles at the limiter/divertor target plates is desired. Peak heat flux is one of the limiting factors in determining the survival probability of plasma-facing-components at high power levels. Plate particle flux affects the neutral flux to the pump, which determines the particle exhaust rate. A technique which couples a two-dimensional (2-D) plasma and a 2-D neutral transport code has been developed (coupled code technique), but this procedure requires large amounts of computer time. Relevant physics has been added to an existing two-neutral-species model which takes the SOL plasma/neutral coupling into account in a simple manner (molecular physics model), and this model is compared with the coupled code technique mentioned above. The molecular physics model is benchmarked against experimental data from a divertor tokamak (DIII-D), and a similar model (single-species model) is benchmarked against data from a pump-limiter tokamak (Tore Supra). The models are then used to examine two key issues: free-streaming-limits (ion energy conduction and momentum flux) and the effects of the non-orthogonal geometry of magnetic flux surfaces and target plates on edge plasma parameter profiles.
Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST
NASA Astrophysics Data System (ADS)
Xu, X. Q.
2008-07-01
We present gyrokinetic neoclassical simulations of tokamak plasmas with a self-consistent electric field 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 backward 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 the gyrokinetic Poisson equation with self-consistent poloidal variation. With a four-dimensional (ψ,θ,γ,μ) version of the TEMPEST code, we compute the radial particle and heat fluxes, the geodesic-acoustic mode, and the development of the neoclassical electric field, which we compare with neoclassical theory using a Lorentz collision model. The present work provides a numerical scheme for self-consistently studying important dynamical aspects of neoclassical transport and electric field in toroidal magnetic fusion devices.
Material Surface Characteristics and Plasma Performance in the Lithium Tokamak Experiment
NASA Astrophysics Data System (ADS)
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
Resistive MHD studies of high-beta Tokamak plasmas
NASA Astrophysics Data System (ADS)
Lynch, V. E.; Hicks, H. R.; Holmes, J. A.; Carreras, B. A.; Garcia, L.
1982-02-01
The magnetohydrodynamic (MHD) activity in high beta Tokamaks such as ISX-B was calculated. These initial value calculations are built on earlier low beta techniques, but the beta effects create several new numerical issues. In addition to time stepping modules, the system of computer codes includes equilibrium solvers (used to provide an initial condition) and output modules, such as a magnetic field line follower and an X-ray diagnostic code. The transition from current driven modes a low beta to predominantly pressure driven modes at high beta is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment.
Edge plasma boundary layer generated by kink modes in tokamaks
Zakharov, Leonid E.
2011-06-15
This paper describes the structure of the electric current generated by external wall touching and free boundary kink modes at the plasma edge using the ideally conducting plasma model. Both kinds of modes generate {delta}-functional surface current at the plasma edge. Free boundary kink modes also perturb the core plasma current, which in the plasma edge compensates the difference between the {delta}-functional surface currents of free boundary and wall touching kink modes. In addition, the resolution of an apparent paradox with the pressure balance across the plasma boundary in the presence of the surface currents is provided.
Edge plasma boundary layer generated by kink modes in tokamaks
NASA Astrophysics Data System (ADS)
Zakharov, Leonid E.
2011-06-01
This paper describes the structure of the electric current generated by external wall touching and free boundary kink modes at the plasma edge using the ideally conducting plasma model. Both kinds of modes generate δ-functional surface current at the plasma edge. Free boundary kink modes also perturb the core plasma current, which in the plasma edge compensates the difference between the δ-functional surface currents of free boundary and wall touching kink modes. In addition, the resolution of an apparent paradox with the pressure balance across the plasma boundary in the presence of the surface currents is provided.
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{sub e} {approx} 1 {minus} 5 {times} 10{sup 14} (cm{sup {minus}3}), T{sub e} {approx} 4 {minus} 10 (eV), B{sub t} {approx} 0.2 {minus} 0.4(T)). Carbon impurity light, mainly the strong lines of C{sub II}(4267A, emitted by the C{sup +} ions) and C{sub III} (4647A, emitted by the C{sup ++} 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{sub {alpha}} 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 {tau}{sub p} is comparable with the plasma energy confinement time {tau}{sub 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.
NASA Astrophysics Data System (ADS)
Shaing, K. C.; Sabbagh, S. A.
2016-07-01
Theory for neoclassical toroidal plasma viscosity has been developed to model transport phenomena, especially, toroidal plasma rotation for tokamaks with broken symmetry. Theoretical predictions are in agreement with the results of the numerical codes in the large aspect ratio limit. The theory has since been extended to include effects of finite aspect ratio and finite plasma β. Here, β is the ratio of the plasma thermal pressure to the magnetic field pressure. However, there are cases where the radial wavelength of the self-consistent perturbed magnetic field strength B on the perturbed magnetic surface is comparable to the width of the trapped particles, i.e., bananas. To accommodate those cases, the theory for neoclassical toroidal plasma viscosity is further extended here to include the effects of the finite banana width. The extended theory is developed using the orbit averaged drift kinetic equation in the low collisionality regimes. The results of the theory can now be used to model plasma transport, including toroidal plasma rotation, in real finite aspect ratio, and finite plasma β tokamaks with the radial wavelength of the perturbed symmetry breaking magnetic field strength comparable to or longer than the banana width.
Scrape-off layer plasma modeling for the DIII-D tokamak
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.
Measurements and modelling of plasma response field to RMP on the COMPASS tokamak
NASA Astrophysics Data System (ADS)
Markovic, T.; Liu, Y. Q.; Cahyna, P.; Pánek, R.; Peterka, M.; Aftanas, M.; Bílková, P.; Bohm, P.; Imríšek, M.; Háček, P.; Havlicek, J.; Havránek, A.; Komm, M.; Urban, J.; Weinzettl, V.; the COMPASS Team
2016-09-01
It has been shown on several tokamaks that application of a resonant magnetic perturbation (RMP) field to the plasma can lead to suppression or mitigation of edge-localized mode (ELM) instabilities. Due to the rotation of the plasma in the RMP field reference system, currents are induced on resonant surfaces within the plasma, consequently screening the original perturbation. In this work, the extensive set of 104 saddle loops installed on the COMPASS tokamak is utilized to measure the plasma response field for two n = 2 RMP configurations of different poloidal mode m spectra. It is shown that spatially the response field is in opposite phase to the original perturbation, and that the poloidal profile of the measured response field does not depend on the poloidal profile of the applied RMP. Simulations of the plasma response by the linear MHD code MARS-F (Liu et al 2000 Phys. Plasmas 7 3681) reveal that both of the studied RMP configurations are well screened by the plasma. Comparison of measured plasma response field with the simulated one shows a good agreement across the majority of poloidal angles, with the exception of the midplane low-field side area, where discrepancy is seen.
Millimeter-Wave Imaging Technology Advancements for Plasma Diagnostics Applications
NASA Astrophysics Data System (ADS)
Kong, Xiangyu
To realize fusion plant, the very first step is to understand the fundamental physics of materials under fusion conditions, i.e. to understand fusion plasmas. Our research group, Plasma Diagnostics Group, focuses on developing advanced tools for physicists to extract as much information as possible from fusion plasmas at millions degrees. The Electron Cyclotron Emission Imaging (ECEI) diagnostics is a very useful tool invented in this group to study fusion plasma electron temperature and it fluctuations. This dissertation presents millimeter wave imaging technology advances recently developed in this group to improve the ECEI system. New technologies made it more powerful to image and visualize magneto-hydrodynamics (MHD) activities and micro-turbulence in fusion plasmas. Topics of particular emphasis start from development of miniaturized elliptical substrate lens array. This novel substrate lens array replaces the previous generation substrate lens, hyper-hemispherical substrate lens, in terms of geometry. From the optical performance perspective, this substitution not only significantly simplifies the optical system with improved optical coupling, but also enhances the RF/LO coupling efficiency. By the benefit of the mini lens focusing properties, a wideband dual-dipole antenna array is carefully designed and developed. The new antenna array is optimized simultaneously for receiving both RF and LO, with sharp radiation patterns, low side-lobe levels, and less crosstalk between adjacent antennas. In addition, a high frequency antenna is also developed, which extends the frequency limit from 145 GHz to 220 GHz. This type of antenna will be used on high field operation tokamaks with toroidal fields in excess of 3 Tesla. Another important technology advance is so-called extended bandwidth double down-conversion electronics. This new electronics extends the instantaneous IF coverage from 2 to 9.2 GHz to 2 to 16.4 GHz. From the plasma point of view, it means that the
Images of plasma disruption effects in the Tokamak Fusion Test Reactor
Maqueda, R.J.; Wurden, G.A.
1999-02-01
Fast-framing imaging of visible radiation from magnetically confined plasmas has lately become a useful tool for both machine operation and physics studies. Using an intensified, commercial Kodak Ektapro imaging system, the effects of a plasma disruption were observed in the Tokamak Fusion Test Reactor (TFTR). The high-energy runaway electrons created soon after the disruption collide with the plasma facing components damaging this surface and producing a shower of debris that traverses the toroidal vessel and falls over the inner bumper limiter.
Observation of spontaneous toroidal rotation inversion in Ohmically heated Tokamak plasmas.
Bortolon, A; Duval, B P; Pochelon, A; Scarabosio, A
2006-12-01
Bulk plasma toroidal rotation is observed to invert spontaneously from counter to cocurrent direction in TCV (Tokamak à Configuration Variable) Ohmically heated discharges, in low confinement mode, without momentum input. The inversion occurs in high current discharges, when the plasma electron density exceeds a well-defined threshold. The transition between the two rotational regimes has been studied by means of density ramps. The results provide evidence of a change of the balance of nondiffusive momentum fluxes in the core of a plasma without an external drive. PMID:17280210
Neoclassical ion thermal conductivity modified by finite banana effects in a tokamak plasma
Chang, C.S.
1997-06-01
A finite-banana-width correction to the neoclassical ion thermal conductivity is obtained in a tokamak plasma under the conventional assumption that the particle flow parallel to magnetic-field lines dominates the trapped particle{close_quote}s orbital dynamics. It is found that the finite-banana-width effect makes ion thermal conductivity itself be a function of radial plasma density gradient and magnetic shear. Negative radial gradients in plasma density and/or safety factor can reduce the neoclassical ion thermal conductivity when the banana width is a significant fraction of the gradient scale length. {copyright} {ital 1997 American Institute of Physics.}
Experimental evidence for self-organized criticality in tokamak plasma turbulence
NASA Astrophysics Data System (ADS)
Rhodes, T. L.; Moyer, R. A.; Groebner, R.; Doyle, E. J.; Lehmer, R.; Peebles, W. A.; Rettig, C. L.
1999-03-01
Measurements of plasma turbulence spectra and particle flux from the DIII-D tokamak exhibit significant agreement with predictions of self-organized criticality (SOC) modeling. Power spectra of density ñ, potential g˜f, and particle flux Γ, are observed to have three regions of frequency dependence: f0, f-1 and f-4. In addition, the particle flux probability distribution displays a Γ-1 scaling over two decades in Γ. These results provide the first evidence that the plasma is in a state consistent with SOC models and place a constraint on plasma transport models.
42GHz ECRH assisted Plasma Breakdown in tokamak SST-1
NASA Astrophysics Data System (ADS)
Shukla, B. K.; Pradhan, S.; Patel, Paresh; Babu, Rajan; Patel, Jatin; Patel, Harshida; Dhorajia, Pragnesh; Tanna, V.; Atrey, P. K.; Manchanda, R.; Gupta, Manoj; Joisa, Shankar; Gupta, C. N.; Danial, Raju; Singh, Prashant; Jha, R.; Bora, D.
2015-03-01
In SST-1, 42GHz ECRH system has been commissioned to carry out breakdown and heating experiments at 0.75T and 1.5T operating toroidal magnetic fields. The 42GHz ECRH system consists of high power microwave source Gyrotron capable to deliver 500kW microwave power for 500ms duration, approximately 20 meter long transmission line and a mirror based launcher. The ECRH power in fundamental O-mode & second harmonic X-mode is launched from low field side (radial port) of the tokamak. At 0.75T operation, approximately 300 kW ECH power is launched in second harmonic X-mode and successful ECRH assisted breakdown is achieved at low loop_voltage ~ 3V. The ECRH power is launched around 45ms prior to loop voltage. The hydrogen pressure in tokamak is maintained ~ 1×10-5mbar and the pre-ionized density is ~ 4×1012/cc. At 1.5T operating toroidal magnetic field, the ECH power is launched in fundamental O-mode. The ECH power at fundamental harmonic is varied from 100 kW to 250 kW and successful breakdown is achieved in all ECRH shots. In fundamental harmonic there is no delay in breakdown while at second harmonic ~ 40ms delay is observed, which is normal in case of second harmonic ECRH assisted breakdown.
NASA Astrophysics Data System (ADS)
Noori, E.; Sadeghi, Y.; Mehdian, H.
2016-06-01
Determination of plasma equilibrium parameters such as poloidal beta (βp) with half of plasma internal inductance (li) known as Shafranov parameter (asymmetry factor) (βp+𝔡li2) and edge safety factor plays very important role in primary equilibrium and stability analysis and control of tokamak plasma. In this study, the well known Shafranov semi-empirical model, based on external magnetic measurements is used to extract Shafranov parameter and effective edge safety factor in low-β operating regime of Damavand tokamak. The well known integral representation of βp+𝔡li2 was modified for non-circular tokamaks with ellipse-like cross section. After calibration of magnetic pick-up coils, Shafranov parameter was estimated with respect to the first and second Fourier harmonic of radial and poloidal components of magnetic field. The results were compared with approximate, semi-analytical determination of Shafranov parameter which is based on analytical solution of Grad-Shafranov equation (GSE). Founding evolution of Shafranov parameter, effective edge safety factor was obtained in terms of Shafranov parameter and compared with semi-empirical description. It was found that between the ramp-up and ramp-down domain of the plasma current, the result from Shafranov model is approximately in good agreement with the semi-analytical and semi-empirical benchmarks and the integral model provides more reliable trace of the Shafranov parameter in out of ramp domains of the discharge.
Kim, Kimin; Ahn, J-W; Scotti, F.; Park, J-K; Menard, J. E.
2015-09-03
Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifies the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.
Kim, Kimin; Ahn, J. -W.; Scotti, F.; Park, J. -K.; Menard, J. E.
2015-09-03
Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifies the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Furthermore, amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.
Long-term development of elongated tokamak plasmas after failure of feedback stabilization
Jensen, T. H.; Chu, M. S.
1989-07-01
Tokamaks with cross sections elongated in the axial direction are subject toan instability that basically involves an axial displacement of the plasma. Thisinstability can be stabilized by a feedback circuit. This Brief Communicationdeals with the long-term development of the plasma after a failure of thefeedback circuit during an otherwise normal discharge. The short-term (linear)development of such a plasma was considered earlier by Yokomizo /ital et/l. (Nucl. Fusion /bold 23/, 1593 (/bold 1983/)). During thelong-term development, the plasma makes intimate contact with the surroundingconducting shell (vacuum vessel). The timescale for this development is longcompared to the Alfven time, so that it is appropriate to consider the plasmain a magnetohydrodynamic (MHD) equlibrium which evolves slowly. This equilibrium(assumed axisymmetric) is unusual in that currents and forces are exchangedbetween the shell and the plasma. The dynamics of the equilibrium is determinedby Ohmic dissipation associated with the plasma currents and the toroidal andpoloidal currents of the shell. For a ''typical'' large tokamak, it is foundthat dissipation in the shell may dominate. Modeling of such long-termdevelopment may be important because the forces acting on the shell due to theshell currents may be large. It may be important that dissipation in the plasmacan be neglected or that it is small in such modeling since the uncertainty ofthe results, due to uncertainties associated with the plasma, is absent or maybe small.
Soft x-ray imaging system for measurement of noncircular tokamak plasmas
Fonck, R.J.; Reusch, M.; Jaehnig, K.P.; Hulse, R.; Roney, P.
1986-08-01
A soft x-ray camera and image processing system has been constructed to provide measurements of the internal shape of high temperature tokamak plasmas. The camera consists of a metallic-foil-filtered pinhole aperture and a microchannel plate image intensifier/convertor which produces a visible image for detection by a CCD TV camera. A wide-angle tangential view of the toroidal plasma allows a single compact camera to view the entire plasma cross section. With Be filters 12 to 50 ..mu..m thick, the signal from the microchannel plate is produced mostly by nickel L-line emissions which orignate in the hot plasma core. The measured toroidal image is numerically inverted to produce a cross-sectional soft x-ray image of the plasma. Since the internal magnetic flux surfaces are usually isothermal and the nickel emissivity depends strongly on the local electron temperature, the x-ray emission contours reflect the shape of the magnetic surfaces in the plasma interior. Initial results from the PBX tokamak experiment show clear differences in internal plasma shapes for circular and bean-shaped discharges.
Importance of Plasma Response to Non-axisymmetric Perturbations in Tokamaks
Jong-kyu Park, Allen H. Boozer, Jonathan E. Menard, Andrea M. Garofalo, Michael J. Schaffer, Richard J. Hawryluk, Stanley M. Kaye, Stefan P. Gerhardt, Steve A. Sabbagh, and the NSTX Team
2009-04-22
Tokamaks are sensitive to deviations from axisymmetry as small as δB=B0 ~ 10-4. These non-axisymmetric perturbations greatly modify plasma confinement and performance by either destroying magnetic surfaces with subsequent locking or deforming magnetic surfaces with associated non-ambipolar transport. The Ideal Perturbed Equilibrium Code (IPEC) calculates ideal perturbed equilibria and provides important basis for understanding the sensitivity of tokamak plasmas to perturbations. IPEC calculations indicate that the ideal plasma response, or equiva- lently the effect by ideally perturbed plasma currents, is essential to explain locking experiments on National Spherical Torus eXperiment (NSTX) and DIII-D. The ideal plasma response is also important for Neoclassical Toroidal Viscosity (NTV) in non-ambipolar transport. The consistency between NTV theory and magnetic braking experiments on NSTX and DIII-D can be improved when the variation in the field strength in IPEC is coupled with generalized NTV theory. These plasma response effects will be compared with the previous vacuum superpositions to illustrate the importance. However, plasma response based on ideal perturbed equilibria is still not suffciently accurate to predict the details of NTV transport, and can be inconsistent when currents associated with a toroidal torque become comparable to ideal perturbed currents.
Rodrigues, Paulo; Bizarro, João P S
2007-09-21
For the first time, tokamak equilibria with negative toroidal current flowing in the plasma core are computed consistently with available measurements from typical current-hole discharges. The equilibrium reconstruction, which leads to non-nested configurations where a system of axisymmetric magnetic islands unfolds, yields an overall good agreement between the computed and experimental plasma-pressure profiles, together with an excellent fit to motional-Stark-effect data. Therefore, considering the accuracy limits of present-day experimental results, care must be exercised when ruling out the existence of tokamak equilibria with central toroidal-current reversal, particularly if relying on reconstruction tools that cannot cope with non-nested configurations. PMID:17930511
GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography
NASA Astrophysics Data System (ADS)
Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.
2015-09-01
An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.
Transport properties of interacting magnetic islands in tokamak plasmas
Gianakon, T.A.; Callen, J.D.; Hegna, C.C.
1993-10-01
This paper explores the equilibrium and transient transport properties of a mixed magnetic topology model for tokamak equilibria. The magnetic topology is composed of a discrete set of mostly non-overlapping magnetic islands centered on the low-order rational surfaces. Transport across the island regions is fast due to parallel transport along the stochastic magnetic field lines about the separatrix of each island. Transport between island regions is assumed to be slow due to a low residual cross-field transport. In equilibrium, such a model leads to: a nonlinear dependence of the heat flux on the pressure gradient; a power balance diffusion coefficient which increases from core to edge; and profile resiliency. Transiently, such a model also exhibits a heat pulse diffusion coefficient larger than the power balance diffusion coefficient.
Resistive MHD studies of high-. beta. -tokamak plasmas
Lynch, V.E.; Carreras, B.A.; Hicks, H.R.; Holmes, J.A.; Garcia, L.
1981-01-01
Numerical calculations have been performed to study the MHD activity in high-..beta.. tokamaks such as ISX-B. These initial value calculations built on earlier low ..beta.. techniques, but the ..beta.. effects create several new numerical issues. These issues are discussed and resolved. In addition to time-stepping modules, our system of computer codes includes equilibrium solvers (used to provide an initial condition) and output modules, such as a magnetic field line follower and an X-ray diagnostic code. The transition from current driven modes at low ..beta.. to predominantly pressure driven modes at high ..beta.. is described. The nonlinear studies yield X-ray emissivity plots which are compared with experiment.
Strong Scattering of High Power Millimeter Waves in Tokamak Plasmas with Tearing Modes
NASA Astrophysics Data System (ADS)
Westerhof, E.; Nielsen, S. K.; Oosterbeek, J. W.; Salewski, M.; de Baar, M. R.; Bongers, W. A.; Bürger, A.; Hennen, B. A.; Korsholm, S. B.; Leipold, F.; Moseev, D.; Stejner, M.; Thoen, D. J.
2009-09-01
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 heating beam. The density determines the detailed phasing of the scattered radiation relative to the O-point passage. The scattering power depends strongly nonlinearly on the heating beam power.
Federici, G.; Skinner, C.H.; Brooks, J.N.; Coad, J.P.; Grisolia, C.
2001-01-10
The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.
Plasma-materials interactions during rf experiments in tokamaks
Cohen, S.A.; Bernabei, S.; Budny, R.; Chu, T.K.; Colestock, P.; Hinnov, E.; Hooke, W.; Hosea, J.; Hwang, D.; Jobes, F.
1984-09-01
Plasma-materials interactions studied in recent ICRF heating and lower hybrid current drive experiments are reviewed. The microscopic processes responsible for impurity generation are discussed. In ICRF experiments, improvements in machine operation and in antenna and feedthrough design have allowed efficient plasma heating at RF powers up to 3 MW. No significant loss of energy from the plasma core due to impurity radiation occurs. Lower hybrid current drive results in the generation and maintenance of hundreds of kiloamperes of plasma current carried by suprathermal electrons. The loss of these electrons and their role in impurity generation are assessed. Methods to avoid this problem are evaluated.
ADVANCES IN COMPREHENSIVE GYROKINETIC SIMULATIONS OF TRANSPORT IN TOKAMAKS
WALTZ,R.E; CANDY,J; HINTON,F.L; ESTRADA-MILA,C; KINSEY,J.E
2004-10-01
A continuum global gyrokinetic code GYRO has been developed to comprehensively simulate core turbulent transport in actual experimental profiles and enable direct quantitative comparisons to the experimental transport flows. GYRO not only treats the now standard ion temperature gradient (ITG) mode turbulence, but also treats trapped and passing electrons with collisions and finite {beta}, equilibrium ExB shear stabilization, and all in real tokamak geometry. Most importantly the code operates at finite relative gyroradius ({rho}{sub *}) so as to treat the profile shear stabilization and nonlocal effects which can break gyroBohm scaling. The code operates in either a cyclic flux-tube limit (which allows only gyroBohm scaling) or globally with physical profile variation. Bohm scaling of DIII-D L-mode has been simulated with power flows matching experiment within error bars on the ion temperature gradient. Mechanisms for broken gyroBohm scaling, neoclassical ion flows embedded in turbulence, turbulent dynamos and profile corrugations, are illustrated.
Advanced diagnostics for plasma chemistry
Kruger, C.H.
1994-03-01
Since July 15, 1992, the High Temperature Gasdynamics Laboratory in the Department of Mechanical Engineering at Stanford University has been engaged in a four-year research program on Advanced Diagnostics for Plasma Chemistry. The goal of this program is to develop state-of-the-art laser-based diagnostics of molecular species in harsh chemical environments, particularly those encountered in plasma synthesis of new materials. Emphasis has been placed on exploiting a new nonlinear spectroscopy, degenerate four wave mixing, as well as linear laser induced fluorescence to accomplish these goals. The present submittal is a proposal for the continuation funding for the third year of this program, from July 15, 1994, until July 14, 1995. Section 2 summarizes the research accomplished during the first eighteen months of the program. Section 3 discusses the plans for continuing research activities. Publications and presentations to date resulting from this program are listed in Section 4. The proposed budget for the third year is given in Section 5.
Liu, X.; Zhao, H. L.; Liu, Y. Li, E. Z.; Han, X.; Ti, A.; Hu, L. Q.; Zhang, X. D.; Domier, C. W.; Luhmann, N. C.
2014-09-15
This paper presents the results of the in situ absolute intensity calibration for the 32-channel heterodyne radiometer on the experimental advanced superconducting tokamak. The hot/cold load method is adopted, and the coherent averaging technique is employed to improve the signal to noise ratio. Measured spectra and electron temperature profiles are compared with those from an independent calibrated Michelson interferometer, and there is a relatively good agreement between the results from the two different systems.
Liu, X; Zhao, H L; Liu, Y; Li, E Z; Han, X; Domier, C W; Luhmann, N C; Ti, A; Hu, L Q; Zhang, X D
2014-09-01
This paper presents the results of the in situ absolute intensity calibration for the 32-channel heterodyne radiometer on the experimental advanced superconducting tokamak. The hot/cold load method is adopted, and the coherent averaging technique is employed to improve the signal to noise ratio. Measured spectra and electron temperature profiles are compared with those from an independent calibrated Michelson interferometer, and there is a relatively good agreement between the results from the two different systems. PMID:25273727
Han, X.; Liu, X.; Liu, Y. Li, E. Z.; Hu, L. Q.; Gao, X.; Domier, C. W.; Luhmann, N. C.
2014-07-15
A 32-channel heterodyne radiometer has been developed for the measurement of electron cyclotron emission (ECE) on the experimental advanced superconducting tokamak (EAST). This system collects X-mode ECE radiation spanning a frequency range of 104–168 GHz, where the frequency coverage corresponds to a full radial coverage for the case with a toroidal magnetic field of 2.3 T. The frequency range is equally spaced every 2 GHz from 105.1 to 167.1 GHz with an RF bandwidth of ∼500 MHz and the video bandwidth can be switched among 50, 100, 200, and 400 kHz. Design objectives and characterization of the system are presented in this paper. Preliminary results for plasma operation are also presented.
NASA Astrophysics Data System (ADS)
Lee, W.; Park, H. K.; Lee, D. J.; Nam, Y. U.; Leem, J.; Kim, T. K.
2016-04-01
The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm-1. The upper limit corresponds to the normalized wavenumber kθρe of ˜0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed.
Han, X. Zhang, T.; Zhang, S. B.; Wang, Y. M.; Shi, T. H.; Liu, Z. X.; Kong, D. F.; Qu, H.; Gao, X.
2014-10-15
Two different pedestal turbulence structures have been observed in edge localized mode-free phase of H-mode heated by lower hybrid wave and RF wave in ion cyclotron range of frequencies (ICRF) on experimental advanced superconducting tokamak. When the fraction of ICRF power P{sub ICRF}/P{sub total} exceeds 0.7, coherent mode is observed. The mode is identified as an electromagnetic mode, rotating in electron diamagnetic direction with a frequency around 50 kHz and toroidal mode number n = −3. Whereas when P{sub ICRF}/P{sub total} is less than 0.7, harmonic mode with frequency f = 40–300 kHz appears instead. The characteristics of these two modes are demonstrated preliminarily. The threshold value of heating power and also the plasma parameters are distinct.
Han, X; Liu, X; Liu, Y; Domier, C W; Luhmann, N C; Li, E Z; Hu, L Q; Gao, X
2014-07-01
A 32-channel heterodyne radiometer has been developed for the measurement of electron cyclotron emission (ECE) on the experimental advanced superconducting tokamak (EAST). This system collects X-mode ECE radiation spanning a frequency range of 104-168 GHz, where the frequency coverage corresponds to a full radial coverage for the case with a toroidal magnetic field of 2.3 T. The frequency range is equally spaced every 2 GHz from 105.1 to 167.1 GHz with an RF bandwidth of ~500 MHz and the video bandwidth can be switched among 50, 100, 200, and 400 kHz. Design objectives and characterization of the system are presented in this paper. Preliminary results for plasma operation are also presented. PMID:25085139
Lee, W; Park, H K; Lee, D J; Nam, Y U; Leem, J; Kim, T K
2016-04-01
The design characteristics of a multi-channel collective (or coherent) scattering system for small scale turbulence study in Korea Superconducting Tokamak Advanced Research (KSTAR), which is planned to be installed in 2017, are given in this paper. A few critical issues are discussed in depth such as the Faraday and Cotton-Mouton effects on the beam polarization, radial spatial resolution, probe beam frequency, polarization, and power. A proper and feasible optics with the 300 GHz probe beam, which was designed based on these issues, provides a simultaneous measurement of electron density fluctuations at four discrete poloidal wavenumbers up to 24 cm(-1). The upper limit corresponds to the normalized wavenumber kθρe of ∼0.15 in nominal KSTAR plasmas. To detect the scattered beam power and extract phase information, a quadrature detection system consisting of four-channel antenna/detector array and electronics will be employed. PMID:27131668
Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions
Bruhn, M.L.
1988-04-01
Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs.
Malik, M.A.
1988-01-01
There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such as INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism.
Module-type flat-field grazing-incidence spectrographs for large Tokamak (JT-60) plasma diagnosis
NASA Astrophysics Data System (ADS)
Nagata, Hiroshi; Kihara, Naoto; Yamashita, Takaji; Sugie, Tatsuo; Kubo, Hirotaka; Shiho, Makoto
1990-09-01
Module-type flat-field grazing-incidence spectrographs with holographic gratings and multichannel detectors for large TOKAMAK (JT-60) plasma diagnosis are developed. The spectrographs cover the different wavelength regions from 0.5-122 nm, and are set to measure impurity lines in the plasma every 20 ms with space resolution of 7 cm. The flat-field imaging properties with designed wavelength resolution were confirmed, and results of tokamak plasma measurements proved the value of these spectrographs for plasma diagnosis.
Improvement of Plasma Performance with Lithium Wall Conditioning in Aditya Tokamak
NASA Astrophysics Data System (ADS)
B. Chowdhuri, M.; Manchanda, R.; Ghosh, J.; B. Bhatt, S.; Ajai, Kumar; K. Das, B.; A. Jadeja, K.; A. Raijada, P.; Manoj, Kumar; Banerjee, S.; Nilam, Ramaiya; Aniruddh, Mali; Ketan, M. Patel; Vinay, Kumar; Vasu, P.; Bhattacharyay, R.; L. Tanna, R.; Y. Shankara, Joisa; K. Atrey, P.; V. S. Rao, C.; Chenna Reddy, D.; K. Chattopadhyay, P.; Jha, R.; C. Saxena, Y.; Aditya Team
2013-02-01
Lithiumization of the vacuum vessel wall of the Aditya tokamak using a lithium rod exposed to glow discharge cleaning plasma has been done to understand its effect on plasma performance. After the Li-coating, an increment of ~100 eV in plasma electron temperature has been observed in most of the discharges compared to discharges without Li coating, and the shot reproducibility is considerably improved. Detailed studies of impurity behaviour and hydrogen recycling are made in the Li coated discharges by observing spectral lines of hydrogen, carbon, and oxygen in the visible region using optical fiber, an interference filter, and PMT based systems. A large reduction in O I signal (up to ~40% to 50%) and a 20% to 30% decrease of Hα signal indicate significant reduction of wall recycling. Furthermore, VUV emissions from O V and Fe XV monitored by a grazing incidence monochromator also show the reduction. Lower Fe XV emission indicates the declined impurity penetration to the core plasma in the Li coated discharges. Significant increase of the particle and energy confinement times and the reduction of Zeff of the plasma certainly indicate the improved plasma parameters in the Aditya tokamak after lithium wall conditioning.
Yang, J H; Yang, X F; Hu, L Q; Zang, Q; Han, X F; Shao, C Q; Sun, T F; Chen, H; Wang, T F; Li, F J; Hu, A L
2013-08-01
A new wide-angle endoscope for visible light observation on the Experimental Advanced Superconducting Tokamak (EAST) has been recently developed. The head section of the optical system is based on a mirror reflection design that is similar to the International Thermonuclear Experimental Reactor-like wide-angle observation diagnostic on the Joint European Torus. However, the optical system design has been simplified and improved. As a result, the global transmittance of the system is as high as 79.6% in the wavelength range from 380 to 780 nm, and the spatial resolution is <5 mm for the full depth of field (4000 mm). The optical system also has a large relative aperture (1:2.4) and can be applied in high-speed camera diagnostics. As an important diagnostic tool, the optical system has been installed on the HT-7 (Hefei Tokamak-7) for its final experimental campaign, and the experiments confirmed that it can be applied to the investigation of transient processes in plasma, such as ELMy eruptions in H-mode, on EAST. PMID:24007102
Yang, J. H.; Hu, L. Q.; Zang, Q.; Han, X. F.; Shao, C. Q.; Sun, T. F.; Chen, H.; Wang, T. F.; Li, F. J.; Hu, A. L.; Yang, X. F.
2013-08-15
A new wide-angle endoscope for visible light observation on the Experimental Advanced Superconducting Tokamak (EAST) has been recently developed. The head section of the optical system is based on a mirror reflection design that is similar to the International Thermonuclear Experimental Reactor-like wide-angle observation diagnostic on the Joint European Torus. However, the optical system design has been simplified and improved. As a result, the global transmittance of the system is as high as 79.6% in the wavelength range from 380 to 780 nm, and the spatial resolution is <5 mm for the full depth of field (4000 mm). The optical system also has a large relative aperture (1:2.4) and can be applied in high-speed camera diagnostics. As an important diagnostic tool, the optical system has been installed on the HT-7 (Hefei Tokamak-7) for its final experimental campaign, and the experiments confirmed that it can be applied to the investigation of transient processes in plasma, such as ELMy eruptions in H-mode, on EAST.
Lee, H. Y.; Hong, J. H.; Jang, J. H.; Park, J. S.; Choe, Wonho; Hahn, S. H.; Bak, J. G.; Lee, J. H.; Ko, W. H.; Lee, K. D.; Lee, S. H.; Lee, H. H.; Juhn, J.-W.; Kim, H. S.; Yoon, S. W.; Han, H.; Ghim, Y.-C.
2015-12-15
It has been reported that supersonic molecular beam injection (SMBI) is an effective means of edge localized mode (ELM) mitigation. This paper newly reports the changes in the ELM, plasma profiles, and fluctuation characteristics during ELM mitigation by SMBI in Korea Superconducting Tokamak Advanced Research. During the mitigated ELM phase, the ELM frequency increased by a factor of 2–3 and the ELM size, which was estimated from the D{sub α} amplitude, the fractional changes in the plasma-stored energy and the line-averaged electron density, and divertor heat flux during an ELM burst, decreased by a factor of 0.34–0.43. Reductions in the electron and ion temperatures rather than in the electron density were observed during the mitigated ELM phase. In the natural ELM phase, frequency chirping of the plasma fluctuations was observed before the ELM bursts; however, the ELM bursts occurred without changes in the plasma fluctuation frequency in the mitigated ELM phase.
Dynamical programming based turbulence velocimetry for fast visible imaging of tokamak plasma.
Banerjee, Santanu; Zushi, H; Nishino, N; Mishra, K; Onchi, T; Kuzmin, A; Nagashima, Y; Hanada, K; Nakamura, K; Idei, H; Hasegawa, M; Fujisawa, A
2015-03-01
An orthogonal dynamic programming (ODP) based particle image velocimetry (PIV) technique is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer (SOL) of tokamaks. This non-intrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights into plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST plasma to estimate the SOL flow in inboard poloidal null-natural divertor configuration. This technique is also applied to investigate the intricate features of the core of the run-away dominated phase following the injection of a large amount of neutrals in the target Ohmic plasma. Development of the ODP-PIV code and its applicability on actual plasma images is reported. PMID:25832227
Qin, H.; Reiman, A.
1996-09-25
An analytic solution is obtained for free-boundary, high-beta equilibria in large aspect ratio tokamaks with a nearly circular plasma boundary. In the absence of surface currents at the plasma-vacuum interface, the free-boundary equilibrium solution introduces constraints arising from the need to couple to an external vacuum field which is physically realizable with a reasonable set of external field coils. This places a strong constraint on the pressure profiles that are consistent with a given boundary shape at high {epsilon}{beta}{sub p}. The equilibrium solution also provides information on the flux surface topology. The plasma is bounded by a separatrix. Increasing the plasma pressure at fixed total current causes the plasma aperture to decrease in a manner that is described.
Comparison of Tokamak Plasma Turbulence Measurements to Self Organized Criticality Modeling
NASA Astrophysics Data System (ADS)
Rhodes, T. L.; Doyle, E. J.; Peebles, W. A.; Rettig, C. L.; Moyer, R. A.; Lehmer, R.; Groebner, R. J.; Thomas, D. M.
1998-11-01
Measurements of plasma turbulence spectra and particle flux from the DIII-D tokamak exhibit significant agreement with predictions of self organized criticality (SOC) modeling [e.g., B. Carreras et al., Phys. Plasmas 3, 2903 (1996)]. To make this comparison an improved method of obtaining turbulent fluctuation spectra in the plasma frame of reference (i.e., where V_E× B≈ E_r/B=0) was used. Utilizing this method, power spectra of density tilde n (both edge and core), potential tildeφ, and particle flux Γ are observed to have three regions of frequency dependence: f^0, f-1, and f-4. In addition, the particle flux probability distribution displays a Γ-1 scaling over two decades in Γ. These results provide the first evidence that the plasma is in a state consistent with SOC models and place a constraint on plasma transport models.
Dynamical programming based turbulence velocimetry for fast visible imaging of tokamak plasma
NASA Astrophysics Data System (ADS)
Banerjee, Santanu; Zushi, H.; Nishino, N.; Mishra, K.; Onchi, T.; Kuzmin, A.; Nagashima, Y.; Hanada, K.; Nakamura, K.; Idei, H.; Hasegawa, M.; Fujisawa, A.
2015-03-01
An orthogonal dynamic programming (ODP) based particle image velocimetry (PIV) technique is developed to measure the time resolved flow field of the fluctuating structures at the plasma edge and scrape off layer (SOL) of tokamaks. This non-intrusive technique can provide two dimensional velocity fields at high spatial and temporal resolution from a fast framing image sequence and hence can provide better insights into plasma flow as compared to conventional probe measurements. Applicability of the technique is tested with simulated image pairs. Finally, it is applied to tangential fast visible images of QUEST plasma to estimate the SOL flow in inboard poloidal null-natural divertor configuration. This technique is also applied to investigate the intricate features of the core of the run-away dominated phase following the injection of a large amount of neutrals in the target Ohmic plasma. Development of the ODP-PIV code and its applicability on actual plasma images is reported.
Edge Plasma Boundary Layer Generated By Kink Modes in Tokamaks
L.E. Zakharov
2010-11-22
This paper describes the structure of the electric current generated by external kink modes at the plasma edge using the ideally conducting plasma model. It is found that the edge current layer is created by both wall touching and free boundary kink modes. Near marginal stability, the total edge current has a universal expression as a result of partial compensation of the δ-functional surface current by the bulk current at the edge. The resolution of an apparent paradox with the pressure balance across the plasma boundary in the presence of the surface currents is provided.
Plasma shape and position controller design for advance plasma configurations in TCV
NASA Astrophysics Data System (ADS)
Anand, Himank; Coda, Stefano; Felici, Federico; Moret, Jean Marc; Le, Hoang Bao
2015-11-01
The performance and stability of tokamak plasma configurations depend strongly on its shape and position. They play a particularly important role in the stability of global magneto-hydrodynamics (MHD) modes and in heat and particle transport. We report on the controller design of a new generalised plasma shape and position controller for advance plasma configurations, using the linearised plasma model RZIP. The controller design is based on an isoflux control scheme and utilises singular value decomposition (SVD), which provides a natural framework for limiting the controlled parameters to the set with the largest singular values, while respecting the combined poloidal field coil current (PF) limits. It also includes the option of weighting the various observers based on the level of importance for a given plasma configuration. The generalised plasma shape and position control algorithm has been successfully tested off-line for limiter and diverted plasma (single null and snowflake configuration) shapes. The testing and commissioning of the controller will commence in the next TCV experimental campaign.
Deuterium-tritium plasmas in novel regimes in the Tokamak Fusion Test Reactor
Bell, M.G.; Beer, M.; Batha, S.
1997-02-01
Experiments in the Tokamak Fusion Test Reactor (TFTR) have explored several novel regimes of improved tokamak confinement in deuterium-tritium (D-T) plasmas, including plasmas with reduced or reversed magnetic shear in the core and high-current plasmas with increased shear in the outer region (high-l{sub i}). New techniques have also been developed to enhance the confinement in these regimes by modifying the plasma-limiter interaction through in-situ deposition of lithium. In reversed-shear plasmas, transitions to enhanced confinement have been observed at plasma currents up to 2.2 MA (q{sub a} {approx} 4.3), accompanied by the formation of internal transport barriers, where large radial gradients develop in the temperature and density profiles. Experiments have been performed to elucidate the mechanism of the barrier formation and its relationship with the magnetic configuration and with the heating characteristics. The increased stability of high-current, high-l{sub i} plasmas produced by rapid expansion of the minor cross-section, coupled with improvement in the confinement by lithium deposition has enabled the achievement of high fusion power, up to 8.7 MW, with D-T neutral beam heating. The physics of fusion alpha-particle confinement has been investigated in these regimes, including the interactions of the alphas with endogenous plasma instabilities and externally applied waves in the ion cyclotron range of frequencies. In D-T plasmas with q{sub 0} > 1 and weak magnetic shear in the central region, a toroidal Alfven eigenmode instability driven purely by the alpha particles has been observed for the first time. The interactions of energetic ions with ion Bernstein waves produced by mode-conversion from fast waves in mixed-species plasmas have been studied as a possible mechanism for transferring the energy of the alphas to fuel ions.
Effect of alpha drift and instabilities on tokamak plasma edge conditions
Miley, G H; Choi, C K
1983-01-01
As suprathermal fusion products slow down in a Tokamak, their average drift is inward. The effect of this drift on the alpha heating and thermalization profiles is examined. In smaller TFTR-type devices, heating in the outer region can be cut in half. Also, the fusion-product energy-distribution near the plasma edge has a positive slope with increasing energy, representing a possible driving mechanism for micro-instabilities. Another instability that can seriously affect outer plasma conditions and shear Alfven transport of alphas is also considered.
Observations of plasma rotation in the high-beta Tokamak Torus 2
NASA Astrophysics Data System (ADS)
Kostek, C.; Marshall, T. C.
Toroidal and poloidal plasma rotation are measured in a high Beta Tokamak device by studying the Doppler sift of the 4686 A He II line. The toroidal flow motion is in the same direction as the plasma current at an average velocity of 1.6 x 10(6) cm/sec, a small fraction of the ion thermal speed. The poloidal flow follows the ion diamagnetic direction, also at an average speed of 1.6 x 10(6) cm/sec. The toroidal flow is compared with the predictions of neoclassical transport theory in the collisional regime. Mechanisms for the time evolution of the rotation are also examined.
Gyrokinetic full-torus simulations of ohmic tokamak plasmas in circular limiter configuration
NASA Astrophysics Data System (ADS)
Korpilo, T.; Gurchenko, A. D.; Gusakov, E. Z.; Heikkinen, J. A.; Janhunen, S. J.; Kiviniemi, T. P.; Leerink, S.; Niskala, P.; Perevalov, A. A.
2016-06-01
The gyrokinetic full 5D particle distribution code ELMFIRE has been extended to simulate circular tokamak plasmas from the magnetic axis to the limiter scrape-off-layer. The predictive power of the code in the full-torus configuration is tested via its ability to reproduce experimental steady-state profiles in FT-2 ohmic L-mode plasmas. The results show that the experimental profile solution is not reproduced numerically due to the difficulty of obtaining global power balance. This is verified by cross-comparison of ELMFIRE code versions, which shows also the impact of boundary conditions and grid resolution on turbulent transport.
NASA Astrophysics Data System (ADS)
Bieg, Bohdan; Chrzanowski, Janusz; Kravtsov, Yury A.; Orsitto, Francesco
Basic principles and recent findings of quasi-isotropic approximation (QIA) of a geometrical optics method are presented in a compact manner. QIA was developed in 1969 to describe electromagnetic waves in weakly anisotropic media. QIA represents the wave field as a power series in two small parameters, one of which is a traditional geometrical optics parameter, equal to wavelength ratio to plasma characteristic scale, and the other one is the largest component of anisotropy tensor. As a result, "" QIA ideally suits to tokamak polarimetry/interferometry systems in submillimeter range, where plasma manifests properties of weakly anisotropic medium.
Effect of poloidal asymmetry on the impurity density profile in tokamak plasmas
Fueloep, T.; Moradi, S.
2011-03-15
The effect of poloidal asymmetry of impurities on impurity transport driven by electrostatic turbulence in tokamak plasmas is analyzed. It is found that if the density of the impurity ions is poloidally asymmetric then the zero-flux impurity density gradient is significantly reduced and even a sign change in the impurity flux may occur if the asymmetry is sufficiently large. This effect is most effective in low shear plasmas with the impurity density peaking on the inboard side and may be a contributing factor to the observed outward convection of impurities in the presence of radio frequency heating.
Edge Temperature Gradient as Intrinsic Rotation Drive in Alcator C-Mod Tokamak Plasmas
Rice, J. E.; Hughes, J. W.; Podpaly, Y. A.; Reinke, M. L.; Greenwald, M. J.; Hubbard, A. E.; Marmar, E. S.; Whyte, D. G.; Diamond, P. H.; Kosuga, Y.; McDevitt, C. J.; Guercan, Oe. D.; Hahm, T. S.
2011-05-27
Intrinsic rotation has been observed in I-mode plasmas from the C-Mod tokamak, and is found to be similar to that in H mode, both in its edge origin and in the scaling with global pressure. Since both plasmas have similar edge {nabla}T, but completely different edge {nabla}n, it may be concluded that the drive of the intrinsic rotation is the edge {nabla}T rather than {nabla}P. Evidence suggests that the connection between gradients and rotation is the residual stress, and a scaling for the rotation from conversion of free energy to macroscopic flow is calculated.
Advanced methods in global gyrokinetic full f particle simulation of tokamak transport
Ogando, F.; Heikkinen, J. A.; Henriksson, S.; Janhunen, S. J.; Kiviniemi, T. P.; Leerink, S.
2006-11-30
A new full f nonlinear gyrokinetic simulation code, named ELMFIRE, has been developed for simulating transport phenomena in tokamak plasmas. The code is based on a gyrokinetic particle-in-cell algorithm, which can consider electrons and ions jointly or separately, as well as arbitrary impurities. The implicit treatment of the ion polarization drift and the use of full f methods allow for simulations of strongly perturbed plasmas including wide orbit effects, steep gradients and rapid dynamic changes. This article presents in more detail the algorithms incorporated into ELMFIRE, as well as benchmarking comparisons to both neoclassical theory and other codes.Code ELMFIRE calculates plasma dynamics by following the evolution of a number of sample particles. Because of using an stochastic algorithm its results are influenced by statistical noise. The effect of noise on relevant magnitudes is analyzed.Turbulence spectra of FT-2 plasma has been calculated with ELMFIRE, obtaining results consistent with experimental data.
Neoclassical Tearing Mode Analysis in Spherical Tokamak Burning Plasmas
NASA Astrophysics Data System (ADS)
Kurita, Daiki; Yamazaki, Kozo; Arimoto, Hideki; Oishi, Tetsutarou; Shoji, Tatsuo
For stabilization of neoclassical tearing mode (NTM), non-resonant helical field (NRHF) is investigated. The time variation of magnetic island is described by modified Rutherford equation. In this work, plasma parameter change due to NTM is analyzed using 1.5-dimensional transport code TOTAL. In ST plasma, magnetic island at 3/2 mode grows by bootstrap current and the central temperature decreases. If NRHF is added, the effect of bootstrap current decreases and NTM is stabilized.
Computational Investigation of Extended-MHD Effects on Tokamak Plasmas
NASA Astrophysics Data System (ADS)
King, Jacob R.; Kruger, Scott E.
2013-10-01
We present studies with the extended-MHD NIMROD code of the tearing instability and edge-localized modes (ELMs). In our first study we use analytics and computations to examine tearing in a large-guide field with a nonzero pressure gradient where previous results show drift effects are stabilizing [Coppi, PoF (1964)]. Our work finds three new results: (1) At moderately large ion gyroradius the mode rotates at the electron drift velocity and there is no stabilization. (2) With collision-less drift reconnection, computations must also include electron gyroviscosity and advection. And (3) we derive a dispersion relation that exhibits diamagnetic stabilization and describes the transition between the electron-fluid-mediated regime of (1) and the semi-collisional regime [Drake and Lee, PoF (1977)]. Our second study investigates the transition from an ideal- to an extended-MHD model in an ELM unstable tokamak configuration. With the inclusion of a full generalized Ohm's law the growth rate is enhanced at intermediate wave-numbers and cut-off at large wave-numbers by diamagnetic effects consistent with analytics [Hastie et al., PoP (2003)]. Adding ion gyroviscosity to the model is stabilizing at large wave-numbers consistent with recent results [Xu et al., PoP (2013)]. Support provided by US DOE.
Characterization of plasma current quench during disruption in EAST tokamak
NASA Astrophysics Data System (ADS)
Chen, Da-Long; Granetz, Robert; Shen, Biao; Yang, Fei; Qian, Jin-Ping; Xiao, Bing-Jia
2015-02-01
A preliminary analysis of plasma current quenching is presented in this paper based on the disruption database. It demonstrates that 26.8% of discharges have been disrupted in the last 2012 campaign, in addition, the plasma disruptive rate grows with the increase of plasma current. The best-fit linear and instantaneous plasma current quench rate is extracted from the recent EAST disruptions, showing that an 80%-30% interval of the maximum plasma current is well fit for the EAST device. The lowest area-normalized current quench time is 3.33 ms/m2 with the estimated plasma electron temperature being 7.3 eV,˜9.5 eV. In the disruption case the maximum eddy current goes up to 400 kA, and a fraction of currents are respectively driven on the upper and lower outer plate with nearly 100 MPa-200 MPa stress in the leg. Project supported by the National Magnetic Confinement Fusion Science Program of China (Grant Nos. 2014GB103000 and 2013GB102000) and the National Natural Science Foundation of China (Grant No. 11205199 and 11205192).
Ding, B. J.; Qin, Y. L.; Li, W. K.; Li, M. H.; Kong, E. H.; Zhang, L.; Wang, M.; Xu, H. D.; Hu, H. C.; Xu, G. S.; Shan, J. F.; Liu, F. K.; Zhao, Y. P.; Wan, B. N.; Li, J. G.; Group, EAST; Ekedahl, A.; Peysson, Y.; Decker, J.
2011-08-15
Using a 2 MW 2.45 GHz lower hybrid wave (LHW) system installed in experimental advanced superconducting tokamak, we have systematically carried out LHW-plasma coupling and lower hybrid current drive experiments in both divertor (double null and lower single null) and limiter plasma configuration with plasma current (I{sub p}) {approx} 250 kA and central line averaged density (n{sub e}) {approx} 1.0-1.3 x 10{sup 19} m{sup -3} recently. Results show that the reflection coefficient (RC) first is flat up to some distance between plasma and LHW grill, and then increases with the distance. Studies indicate that with the same plasma parameters, the best coupling is obtained in the limiter case (with plasma leaning on the inner wall), followed by the lower single null, and the one with the worst coupling is the double null configuration, explained by different magnetic connection length. The RCs in the different poloidal rows show that they have different coupling characteristics, possibly due to local magnetic connection length. Current drive efficiency has been investigated by a least squares fit with N{sub //}{sup peak}=2.1, where N{sub //}{sup peak} is the peak value of parallel refractive index of the launched wave. Results show that there is no obvious difference in the current drive efficiency between double null and lower single null cases, whereas the efficiency is somewhat small in the limiter configuration. This is in agreement with the ray tracing/Fokker-Planck code simulation by LUKE/C3PO and can be interpreted by the power spectrum up-shift factor in different plasma configurations. A transformer recharge is realized with {approx}0.8 MW LHW power and the energy conversion efficiency from LHW to poloidal field energy is about 2%.
NASA Astrophysics Data System (ADS)
Ding, B. J.; Qin, Y. L.; Li, W. K.; Li, M. H.; Kong, E. H.; Zhang, L.; Ekedahl, A.; Peysson, Y.; Decker, J.; Wang, M.; Xu, H. D.; Hu, H. C.; Xu, G. S.; Shan, J. F.; Liu, F. K.; Zhao, Y. P.; Wan, B. N.; Li, J. G.; Group, EAST
2011-08-01
Using a 2 MW 2.45 GHz lower hybrid wave (LHW) system installed in experimental advanced superconducting tokamak, we have systematically carried out LHW-plasma coupling and lower hybrid current drive experiments in both divertor (double null and lower single null) and limiter plasma configuration with plasma current (Ip) ˜ 250 kA and central line averaged density (ne) ˜ 1.0-1.3 × 1019 m-3 recently. Results show that the reflection coefficient (RC) first is flat up to some distance between plasma and LHW grill, and then increases with the distance. Studies indicate that with the same plasma parameters, the best coupling is obtained in the limiter case (with plasma leaning on the inner wall), followed by the lower single null, and the one with the worst coupling is the double null configuration, explained by different magnetic connection length. The RCs in the different poloidal rows show that they have different coupling characteristics, possibly due to local magnetic connection length. Current drive efficiency has been investigated by a least squares fit with N//peak=2.1, where N//peak is the peak value of parallel refractive index of the launched wave. Results show that there is no obvious difference in the current drive efficiency between double null and lower single null cases, whereas the efficiency is somewhat small in the limiter configuration. This is in agreement with the ray tracing/Fokker-Planck code simulation by LUKE/C3PO and can be interpreted by the power spectrum up-shift factor in different plasma configurations. A transformer recharge is realized with ˜0.8 MW LHW power and the energy conversion efficiency from LHW to poloidal field energy is about 2%.
Cryogenic pellet launcher adapted for controlling of tokamak plasma edge instabilities.
Lang, P T; Cierpka, P; Harhausen, J; Neuhauser, J; Wittmann, C; Gál, K; Kálvin, S; Kocsis, G; Sárközi, J; Szepesi, T; Dorner, C; Kauke, G
2007-02-01
One of the main challenges posed recently on pellet launcher systems in fusion-oriented plasma physics is the control of the plasma edge region. Strong energy bursts ejected from the plasma due to edge localized modes (ELMs) can form a severe threat for in-vessel components but can be mitigated by sufficiently frequent triggering of the underlying instabilities using hydrogen isotope pellet injection. However, pellet injection systems developed mainly for the task of ELM control, keeping the unwanted pellet fueling minimized, are still missing. Here, we report on a novel system developed under the premise of its suitability for control and mitigation of plasma edge instabilities. The system is based on the blower gun principle and is capable of combining high repetition rates up to 143 Hz with low pellet velocities. Thus, the flexibility of the accessible injection geometry can be maximized and the pellet size kept low. As a result the new system allows for an enhancement in the tokamak operation as well as for more sophisticated experiments investigating the underlying physics of the plasma edge instabilities. This article reports on the design of the new system, its main operational characteristics as determined in extensive test bed runs, and also its first test at the tokamak experiment ASDEX Upgrade. PMID:17578110
Cryogenic pellet launcher adapted for controlling of tokamak plasma edge instabilities
Lang, P. T.; Cierpka, P.; Harhausen, J.; Neuhauser, J.; Wittmann, C.; Gal, K.; Kalvin, S.; Kocsis, G.; Sarkoezi, J.; Szepesi, T.; Dorner, C.; Kauke, G.
2007-02-15
One of the main challenges posed recently on pellet launcher systems in fusion-oriented plasma physics is the control of the plasma edge region. Strong energy bursts ejected from the plasma due to edge localized modes (ELMs) can form a severe threat for in-vessel components but can be mitigated by sufficiently frequent triggering of the underlying instabilities using hydrogen isotope pellet injection. However, pellet injection systems developed mainly for the task of ELM control, keeping the unwanted pellet fueling minimized, are still missing. Here, we report on a novel system developed under the premise of its suitability for control and mitigation of plasma edge instabilities. The system is based on the blower gun principle and is capable of combining high repetition rates up to 143 Hz with low pellet velocities. Thus, the flexibility of the accessible injection geometry can be maximized and the pellet size kept low. As a result the new system allows for an enhancement in the tokamak operation as well as for more sophisticated experiments investigating the underlying physics of the plasma edge instabilities. This article reports on the design of the new system, its main operational characteristics as determined in extensive test bed runs, and also its first test at the tokamak experiment ASDEX Upgrade.
Gorini, G.; Kaellne, J.; Ognissanto, F.; Tardocchi, M.
2011-03-15
A parametric relationship between total neutron yield rate and collimated fluxes related to the brightness (B) of plasma chords ({lambda}) is developed for different emissivity distributions of tokamak plasmas. Specifically, the brightness was expressed as a function of chord coordinates of radial position using a simple model for the emissivity profiles of width parameter w. The functional brightness dependence B({lambda},w) was calculated to examine the relationship between measured flux and deduced yield rate, and its plasma profile dependence. The results were used to determine the chord range of minimum profile sensitivity in order to identify the preferred collimator sight for the determination of yield rate from neutron emission spectroscopy (YNES) measurements. The YNES method is discussed in comparison to conventional methods to determine the total neutron yield rates and related plasma fusion power relying on uncollimated flux measurements and a different calibration base for the flux-yield relationship. The results have a special bearing for tokamaks operating with both deuterium and deuterium-tritium plasmas and future high power machines such as for ITER, DEMO, and IGNITOR.
Phase locking of multi-helicity neoclassical tearing modes in tokamak plasmas
NASA Astrophysics Data System (ADS)
Fitzpatrick, Richard
2015-04-01
The attractive "hybrid" tokamak scenario combines comparatively high q95 operation with improved confinement compared with the conventional H98 ,y 2 scaling law. Somewhat unusually, hybrid discharges often exhibit multiple neoclassical tearing modes (NTMs) possessing different mode numbers. The various NTMs are eventually observed to phase lock to one another, giving rise to a significant flattening, or even an inversion, of the core toroidal plasma rotation profile. This behavior is highly undesirable because the loss of core plasma rotation is known to have a deleterious effect on plasma stability. This paper presents a simple, single-fluid, cylindrical model of the phase locking of two NTMs with different poloidal and toroidal mode numbers in a tokamak plasma. Such locking takes place via a combination of nonlinear three-wave coupling and conventional toroidal coupling. In accordance with experimental observations, the model predicts that there is a bifurcation to a phase-locked state when the frequency mismatch between the modes is reduced to one half of its original value. In further accordance, the phase-locked state is characterized by the permanent alignment of one of the X-points of NTM island chains on the outboard mid-plane of the plasma, and a modified toroidal angular velocity profile, interior to the outermost coupled rational surface, which is such that the core rotation is flattened, or even inverted.
Phase locking of multi-helicity neoclassical tearing modes in tokamak plasmas
Fitzpatrick, Richard
2015-04-15
The attractive “hybrid” tokamak scenario combines comparatively high q{sub 95} operation with improved confinement compared with the conventional H{sub 98,y2} scaling law. Somewhat unusually, hybrid discharges often exhibit multiple neoclassical tearing modes (NTMs) possessing different mode numbers. The various NTMs are eventually observed to phase lock to one another, giving rise to a significant flattening, or even an inversion, of the core toroidal plasma rotation profile. This behavior is highly undesirable because the loss of core plasma rotation is known to have a deleterious effect on plasma stability. This paper presents a simple, single-fluid, cylindrical model of the phase locking of two NTMs with different poloidal and toroidal mode numbers in a tokamak plasma. Such locking takes place via a combination of nonlinear three-wave coupling and conventional toroidal coupling. In accordance with experimental observations, the model predicts that there is a bifurcation to a phase-locked state when the frequency mismatch between the modes is reduced to one half of its original value. In further accordance, the phase-locked state is characterized by the permanent alignment of one of the X-points of NTM island chains on the outboard mid-plane of the plasma, and a modified toroidal angular velocity profile, interior to the outermost coupled rational surface, which is such that the core rotation is flattened, or even inverted.
Fast wave heating and current drive in tokamak plasmas with negative central shear
Forest, C.B.; Petty, C.C.; Baity, F.W.
1996-07-01
Fast waves provide an excellent tool for heating electrons and driving current in the central region of tokamak plasmas. In this paper, we report the use of centrally peaked electron heating and current drive to study transport in plasmas with negative central shear (NCS). Tokamak plasmas with NCS offer the potential of reduced energy transport and improved MHD stability properties, but will require non-inductive current drive to maintain the required current profiles. Fast waves, combined with neutral beam injection, provide the capability to change the central current density evolution and independently vary {ital T{sub e}}, and {ital T{sub i}} for transport studies in these plasmas. Electron heating also reduces the collisional heat exchange between electrons and ions and reduces the power deposition from neutral beams into electrons, thus improving the certainty in the estimate of the electron heating. The first part of this paper analyzes electron and ion heat transport in the L-mode phase of NCS plasmas as the current profile resistively evolves. The second part of the paper discusses the changes that occur in electron as well as ion energy transport in this phase of improved core confinement associated with NCS.
Hu, Shilin; Qu, Hongpeng; Li, Jiquan; Kishimoto, Y.
2014-10-15
Resistive drift wave instability is investigated numerically in tokamak edge plasma confined by sheared slab magnetic field geometry with an embedded magnetic island. The focus is on the structural characteristics of eigenmode inside the island, where the density profile tends to be flattened. A transition of the dominant eigenmode occurs around a critical island width w{sub c}. For thin islands with a width below w{sub c}, two global long wavelength eigenmodes with approximately the same growth rate but different eigenfrequency are excited, which are stabilized by the magnetic island through two-dimensional mode coupling in both x and y (corresponding to radial and poloidal in tokamak) directions. On the other hand, a short wavelength eigenmode, which is destabilized by thick islands with a width above w{sub c}, dominates the edge fluctuation, showing a prominent structural localization in the region between the X-point and the O-point of the magnetic island. The main destabilization mechanism is identified as the mode coupling in the y direction, which is similar to the so-called toroidal coupling in tokamak plasmas. These three eigenmodes may coexist in the drift wave fluctuation for the island with a width around w{sub c}. It is demonstrated that the structural localization results mainly from the quasilinear flattening of density profile inside the magnetic island.
Integrated modelling of the Globus-M tokamak plasma and a comparison with SOL width scaling
NASA Astrophysics Data System (ADS)
Senichenkov, I. Yu.; Kaveeva, E. G.; Gogoleva, A. V.; Vekshina, E. O.; Zadvitskiy, G. V.; Molchanov, P. A.; Rozhansky, V. A.; Voskoboynikov, S. P.; Khromov, N. A.; Lepikhov, S. A.; Gusev, V. K.; The Globus-M Team
2015-05-01
Recently a scheme for the coupling of the one-dimensional core transport code ASTRA and the two-dimensional edge transport code B2SOLPS was developed, thus providing the integrated modelling of tokamak discharge. Here, this scheme is improved by taking impurities into account and by considering a real flux surface shape using the equilibrium code SPIDER. This integrated modelling is applied to discharges of the spherical tokamak Globus-M to study the dependence of the scrape-off layer (SOL) width and divertor heat loads on the discharge power and the plasma current. Since these values, together with the magnetic field, are relatively small in Globus-M, this study can test the existing scaling against data in a wider range of tokamak operational parameters. The modelling results agree reasonably with Thomson scattering and Langmuir probe measurements and allow, in principle, the determination of the physical mechanisms responsible for the SOL structure formation. It is found that the SOL width is approximately inversely proportional to the plasma current, in agreement with existing experimental scaling, while its dependence on discharge power is found to be quite weak.
NASA Astrophysics Data System (ADS)
Frantz, Eric Randall
Elongation and shaping of the tokamak plasma cross -section can allow increased beta and other favorable improvements. As the cross-section is made non-circular, however, the plasma can become unstable against axisymmetric motions, the most predominant one being a nearly uniform displacement in the direction of elongation. Without additional stabilizing mechanisms, this instability has growth rates typically (TURN)10('6)sec('-1). With passive and active feedback from external conductors, the plasma can be significantly slowed down and controlled. In this work, a mathematical formulism for analyzing the vertical instability is developed in which the external conductors are treated (or broken -up) as discrete coils. The circuit equations for the plasma induced currents can be included within the same mathematical framework. The plasma equation of motion and the circuit equations are combined and manipulated into a diagonalized form that can be graphically analyzed to determine the growth rate. An effective mode approximation (EMA) to the dispersion relation in introduced to simplify and approximate the growth rate of the more exact case. Controller voltage equations for active feedback are generalized to include position and velocity feedback and time delay. A position cut-off displacement is added to model finite spatial resolution of the position detectors or a dead-band voltage level. Stability criteria are studied for EMA and the more exact case. The time dependent responses for plasma position controller voltages, and currents are determined from the Laplace transformations. Slow responses are separated from the fast ones (dependent on plasma inertia) using a typical tokamak ordering approximation. The methods developed are applied in numerous examples for the machine geometry and plasma of TNS, an inside-D configuration plasma resembling JET, INTOR, or FED.
Trapped electron mode turbulence driven intrinsic rotation in Tokamak plasmas.
Wang, W X; Hahm, T S; Ethier, S; Zakharov, L E; Diamond, P H
2011-02-25
Progress from global gyrokinetic simulations in understanding the origin of intrinsic rotation in toroidal plasmas is reported. The turbulence-driven intrinsic torque associated with nonlinear residual stress generation due to zonal flow shear induced asymmetry in the parallel wave number spectrum is shown to scale close to linearly with plasma gradients and the inverse of the plasma current, qualitatively reproducing experimental empirical scalings of intrinsic rotation. The origin of current scaling is found to be enhanced k(∥) symmetry breaking induced by the increased radial variation of the safety factor as the current decreases. The intrinsic torque is proportional to the pressure gradient because both turbulence intensity and zonal flow shear, which are two key ingredients for driving residual stress, increase with turbulence drive, which is R/L(T(e)) and R/L(n(e)) for the trapped electron mode. PMID:21405577
A novel local equilibrium model for shaped tokamak plasmas
Yu Weihong; Zhou Deng; Xiang Nong
2012-07-15
A model is proposed for a local up-down symmetric equilibrium in the vicinity of a specified magnetic surface with given elongation and triangularity. Different from the Miller's model [R. L. Miller et al., Phys. Plasmas 5, 973 (1998)], the derivative of the Shafranov shift in the present model is self-consistently determined. The equilibrium accounts for all the essential features, like the elongation, the triangularity, and the Shafranov shift etc., of a shaped cross section. Hence, it can be used for investigation of radially localized plasma modes, like reversed shear Alfvenic eigenmodes and ballooning mode, etc., and it is also suitable for local equilibrium construction used for flux tube plasma simulations.
Natural Divertor Spherical Tokamak Plasmas with bean shape and ergodic limiter
NASA Astrophysics Data System (ADS)
Ribeiro, Celso; Herrera, Julio; Chavez, Esteban; Tritz, Kevin
2013-10-01
The former spherical tokamak (ST) MEDUSA (Madison EDUcation Small Aspect.ratio tokamak, R < 0.14 m, a < 0.10 m, BT < 0.5T, Ip < 40 kA, 3 ms pulse) is being recommissioned in Costa Rica Institute of Technology. The main objectives of the MEDUSA-CR project are training and to clarify several issues in relevant physics for conventional and mainly STs, including beta studies in bean-shaped ST plasmas, transport, heating and current drive via Alfvén wave, and natural divertor STs with ergodic magnetic limiter. We report here improvements in the self-consistency of these equilibrium comparisons and a preliminary study of their MHD stability beta limits. VIE-ITCR, IAEA-CRP contract 17592, National Instruments of Costa Rica.
Cohen, R H; Fielding, S; Helander, P; Ryutov, D D
2001-09-05
This paper surveys theory issues associated with inducing convective cells through divertor tile biasing in a tokamak to broaden the scrape-off layer (SOL). The theory is applied to the Mega-Ampere Spherical Tokamak (MAST), where such experiments are planned in the near future. Criteria are presented for achieving strong broadening and for exciting shear-flow turbulence in the SOL; these criteria are shown to be attainable in practice. It is also shown that the magnetic shear present in the vicinity of the X-point is likely to confine the potential perturbations to the divertor region below the X-point, leaving the part of the SOL that is in direct contact with the core plasma intact. The current created by the biasing and the associated heating power are found to be modest.
x-ray irradiation analysis based on wavelet transform in tokamak plasma.
Ghanbari, K; Ghoranneviss, M; Elahi, A Salar; Saviz, S
2014-01-01
Hard x-ray emission from the Runaway electrons is an important issue in tokamaks. Suggesting methods to reduce the Runaway electrons and therefore the emitted hard x-ray is important for tokamak plasma operation. In this manuscript, we have investigated the effects of external fields on hard x-ray intensity and Magneto-Hydro-Dynamic (MHD) activity. In other words, we have presented the effects of positive biased limiter and Resonant Helical Field (RHF) on the MHD fluctuations and hard x-ray emission from the Runaway electrons. MHD activity and hard x-ray intensity were analyzed using Wavelet transform in the presence of external fields and without them. The results show that the MHD activity and therefore the hard x-ray intensity can be controlled by the external electric and magnetic fields. PMID:25408394
Recent Advancements in Microwave Imaging Plasma Diagnostics
H. Park; C.C. Chang; B.H. Deng; C.W. Domier; A.J.H. Donni; K. Kawahata; C. Liang; X.P. Liang; H.J. Lu; N.C. Luhmann, Jr.; A. Mase; H. Matsuura; E. Mazzucato; A. Miura; K. Mizuno; T. Munsat; K. and Y. Nagayama; M.J. van de Pol; J. Wang; Z.G. Xia; W-K. Zhang
2002-03-26
Significant advances in microwave and millimeter wave technology over the past decade have enabled the development of a new generation of imaging diagnostics for current and envisioned magnetic fusion devices. Prominent among these are revolutionary microwave electron cyclotron emission imaging (ECEI), microwave phase imaging interferometers, imaging microwave scattering and microwave imaging reflectometer (MIR) systems for imaging electron temperature and electron density fluctuations (both turbulent and coherent) and profiles (including transport barriers) on toroidal devices such as tokamaks, spherical tori, and stellarators. The diagnostic technology is reviewed, and typical diagnostic systems are analyzed. Representative experimental results obtained with these novel diagnostic systems are also presented.
Graf, A; May, M; Beiersdorfer, P; Magee, E; Lawrence, M; Terry, J; Rice, J
2004-04-29
We present a high throughput (f/3) visible (3500 - 7000 Angstrom) Doppler spectrometer for toroidal rotation velocity measurements of the Alcator C-Mod tokamak plasma. The spectrometer has a temporal response of 1 ms and a rotation velocity sensitivity of {approx}10{sup 5} cm/s. This diagnostic will have a tangential view and map out the plasma rotation at several locations along the outer half of the minor radius (r/a > 0.5). The plasma rotation will be determined from the Doppler shifted wavelengths of D{sub alpha} and magnetic and electric dipole transitions of highly ionized impurities in the plasma. The fast time resolution and high spectral resolving power are possible due to a 6' diameter circular transmission grating that is capable of {lambda}/{Delta}{lambda} {approx} 15500 at 5769 Angstrom in conjunction with a 50 {micro}m slit.
SUSTAINED STABILIZATION OF THE RESISTIVE WALL MODE BY PLASMA ROTATION IN THE DIII-D TOKAMAK
GAROFALO,A.M; STRAIT,E.J; JOHNSON,L.C; LA HAYE,R.J; LAZARUS,E.A; NAVRATIL,G.A; OKABAYASHI,M; SCOVILLE,J.T; TAYLOR,T.S; TURNBULL,A.D; AND THE DIII-D TEAM
2001-10-01
OAK-B135 A path to sustained stable operation, at plasma pressure up to twice the ideal magnetohydrodynamic (MHD) n = 1 free-boundary pressure limit, has been discovered in the DIII-D tokamak. Tuning the correction of the intrinsic magnetic field asymmetries so as to minimize plasma rotation decay during the high beta phase and increasing the angular momentum injection, have allowed maintaining the plasma rotation above that needed for stabilization of the resistive wall mode (RWM). A new method to determine the improved magnetic field correction uses feedback to sense and minimize the resonant plasma response to the non-axisymmetric field. At twice the free-boundary pressure limit, a disruption precursor is observed, which is consistent with having reached the ''ideal wall'' pressure limit predicted by stability calculations.
The interaction between fishbone modes and shear Alfvén waves in tokamak plasmas
NASA Astrophysics Data System (ADS)
He, Hongda; Liu, Yueqiang; Dong, J. Q.; Hao, G. Z.; Wu, Tingting; He, Zhixiong; Zhao, K.
2016-05-01
The resonant interaction between the energetic particle triggered fishbone mode and the shear Alfvén waves is computationally investigated and firmly demonstrated based on a tokamak plasma equilibrium, using the self-consistent MHD-kinetic hybrid code MARS-K (Liu et al 2008 Phys. Plasmas 15 112503). This type of continuum resonance, occurring critically due to the mode’s toroidal rotation in the plasma frame, significantly modifies the eigenmode structure of the fishbone instability, by introducing two large peaks of the perturbed parallel current density near but offside the q = 1 rational surface (q is the safety factor). The self-consistently computed radial plasma displacement substantially differs from that being assumed in the conventional fishbone theory.
Modeling of transport phenomena in tokamak plasmas with neural networks
Meneghini, O.; Luna, C. J.; Smith, S. P.; Lao, L. L.
2014-06-15
A new transport model that uses neural networks (NNs) to yield electron and ion heat flux profiles has been developed. Given a set of local dimensionless plasma parameters similar to the ones that the highest fidelity models use, the NN model is able to efficiently and accurately predict the ion and electron heat transport profiles. As a benchmark, a NN was built, trained, and tested on data from the 2012 and 2013 DIII-D experimental campaigns. It is found that NN can capture the experimental behavior over the majority of the plasma radius and across a broad range of plasma regimes. Although each radial location is calculated independently from the others, the heat flux profiles are smooth, suggesting that the solution found by the NN is a smooth function of the local input parameters. This result supports the evidence of a well-defined, non-stochastic relationship between the input parameters and the experimentally measured transport fluxes. The numerical efficiency of this method, requiring only a few CPU-μs per data point, makes it ideal for scenario development simulations and real-time plasma control.
NASA Astrophysics Data System (ADS)
Popov, Tsv K.; Dimitrova, M.; Ivanova, P.; Kovačič, J.; Gyergyek, T.; Dejarnac, R.; Stöckel, J.; Pedrosa, M. A.; López-Bruna, D.; Hidalgo, C.
2016-06-01
Advanced Langmuir probe techniques for evaluating the plasma potential and electron-energy distribution function (EEDF) in magnetized plasma are reviewed. It is shown that when the magnetic field applied is very weak and the electrons reach the probe without collisions in the probe sheath the second-derivative Druyvesteyn formula can be used for EEDF evaluation. At low values of the magnetic field, an extended second-derivative Druyvesteyn formula yields reliable results, while at higher values of the magnetic field, the first-derivative probe technique is applicable for precise evaluation of the plasma potential and the EEDF. There is an interval of intermediate values of the magnetic field when both techniques—the extended second-derivative and the first-derivative one—can be used. Experimental results from probe measurements in different ranges of magnetic field are reviewed and discussed: low-pressure argon gas discharges in the presence of a magnetic field in the range from 0.01 to 0.08 T, probe measurements in circular hydrogen plasmas for high-temperature fusion (magnetic fields from 0.45 T to 1.3 T) in small ISTTOK and CASTOR tokamaks, D-shape COMPASS tokamak plasmas, as well as in the TJ-II stellarator. In the vicinity of the last closed flux surface (LCFS) in tokamaks and in the TJ-II stellarator, the EEDF obtained is found to be bi-Maxwellian, while close to the tokamak chamber wall it is Maxwellian. The mechanism of the appearance of a bi-Maxwellian EEDF in the vicinity of the LCFS is discussed. Comparison of the results from probe measurements with those obtained from calculations using the ASTRA and EIRENE codes shows that the main reason for the appearance of a bi-Maxwellian EEDF in the vicinity of the LCFS is the ionization of the neutral atoms.
Scattering of radio frequency waves by blobs in tokamak plasmas
Ram, Abhay K.; Hizanidis, Kyriakos; Kominis, Yannis
2013-05-15
The density fluctuations and blobs present in the edge region of magnetic fusion devices can scatter radio frequency (RF) waves through refraction, reflection, diffraction, and coupling to other plasma waves. This, in turn, affects the spectrum of the RF waves and the electromagnetic power that reaches the core of the plasma. The usual geometric optics analysis of RF scattering by density blobs accounts for only refractive effects. It is valid when the amplitude of the fluctuations is small, of the order of 10%, compared to the background density. In experiments, density fluctuations with much larger amplitudes are routinely observed, so that a more general treatment of the scattering process is needed. In this paper, a full-wave model for the scattering of RF waves by a blob is developed. The full-wave approach extends the range of validity well beyond that of geometric optics; however, it is theoretically and computationally much more challenging. The theoretical procedure, although similar to that followed for the Mie solution of Maxwell's equations, is generalized to plasmas in a magnetic field. Besides diffraction and reflection, the model includes coupling to a different plasma wave than the one imposed by the external antenna structure. In the model, it is assumed that the RF waves interact with a spherical blob. The plasma inside and around the blob is cold, homogeneous, and imbedded in a uniform magnetic field. After formulating the complete analytical theory, the effect of the blob on short wavelength electron cyclotron waves and longer wavelength lower hybrid waves is studied numerically.
ASPECT: An advanced specified-profile evaluation code for tokamaks
Stotler, D.P.; Reiersen, W.T.; Bateman, G.
1993-03-01
A specified-profile, global analysis code has been developed to evaluate the performance of fusion reactor designs. Both steady-state and time-dependent calculations are carried out; the results of the former can be used in defining the parameters of the latter, if desired. In the steady-state analysis, the performance is computed at a density and temperature chosen to be consistent with input limits (e.g., density and beta) of several varieties. The calculation can be made at either the intersection of the two limits or at the point of optimum performance as the density and temperature are varied along the limiting boundaries. Two measures of performance are available for this purpose: the ignition margin or the confinement level required to achieve a prescribed ignition margin. The time-dependent calculation can be configured to yield either the evolution of plasma energy as a function of time or, via an iteration scheme, the amount of auxiliary power required to achieve a desired final plasma energy.
Generation of plasma rotation in a tokamak by ion-cyclotron absorption of fast Alfven waves
F.W. Perkins; R.B. White; P. Bonoli
2000-06-13
Control of rotation in tokamak plasmas provides a method for suppressing fine-scale turbulent transport by velocity shear and for stabilizing large-scale magnetohydrodynamic instabilities via a close-fitting conducting shell. The experimental discovery of rotation in a plasma heated by the fast-wave minority ion cyclotron process is important both as a potential control method for a fusion reactor and as a fundamental issue, because rotation arises even though this heating process introduces negligible angular momentum. This paper proposes and evaluates a mechanism which resolves this apparent conflict. First, it is assumed that angular momentum transport in a tokamak is governed by a diffusion equation with a no-slip boundary condition at the plasma surface and with a torque-density source that is a function of radius. When the torque density source consists of two separated regions of positive and negative torque density, a non-zero central rotation velocity results, even when the total angular momentum input vanishes. Secondly, the authors show that localized ion-cyclotron heating can generate regions of positive and negative torque density and consequently central plasma rotation.
Renewable boron carbide coating in plasma shots of tokamak Т11-М
NASA Astrophysics Data System (ADS)
Buzhinskij, O. I.; Barsuk, V. A.; Otroshchenko, V. G.
2009-06-01
Experimental results on boronization in plasma shots of the tokamak T-11M are presented. Non-toxic and not explosive metacarborane C 2H 12B 10 was used in the boron deposition process. Experiments have been carried out in shots with parameters: toroidal field ˜1-1.2 Т, plasma current Ip = 70 кА, average shot duration tp ˜ 150 ms and electron density along the central chord ne ˜ 2.5 × 10 13 cm -3. As a result of experiment, a dense film of ˜0.2 microns thickness with good adhesion to a surface has formed on the reference specimens after 8 s boronization. After boronization the impurities in wall areas have been suppressed. High vacuum characteristics of the discharge chamber were stabilized. Working vacuum was reached without a preliminary induction heating and cleaning by a glow discharge, and stabilization of the plasma filament has improved. Shot duration without disruption at densities of ne = 1.3 × 10 13 сm -3, Ip = 70 кА was 350 ms and ne = 4.64 × 10 13 сm -3, Ip = 70 кА was 250 ms. High repeatability of experimental results has appeared. Developed technology opens an opportunity of practical production of renewable structured boron-carbon coatings with use of plasma shots in large-scale tokamaks, such as DIII-D, JET, JT-60 UP, ITER, DEMO.
Identification and control of plasma vertical position using neural network in Damavand tokamak
Rasouli, H.; Rasouli, C.; Koohi, A.
2013-02-15
In this work, a nonlinear model is introduced to determine the vertical position of the plasma column in Damavand tokamak. Using this model as a simulator, a nonlinear neural network controller has been designed. In the first stage, the electronic drive and sensory circuits of Damavand tokamak are modified. These circuits can control the vertical position of the plasma column inside the vacuum vessel. Since the vertical position of plasma is an unstable parameter, a direct closed loop system identification algorithm is performed. In the second stage, a nonlinear model is identified for plasma vertical position, based on the multilayer perceptron (MLP) neural network (NN) structure. Estimation of simulator parameters has been performed by back-propagation error algorithm using Levenberg-Marquardt gradient descent optimization technique. The model is verified through simulation of the whole closed loop system using both simulator and actual plant in similar conditions. As the final stage, a MLP neural network controller is designed for simulator model. In the last step, online training is performed to tune the controller parameters. Simulation results justify using of the NN controller for the actual plant.
Identification and control of plasma vertical position using neural network in Damavand tokamak.
Rasouli, H; Rasouli, C; Koohi, A
2013-02-01
In this work, a nonlinear model is introduced to determine the vertical position of the plasma column in Damavand tokamak. Using this model as a simulator, a nonlinear neural network controller has been designed. In the first stage, the electronic drive and sensory circuits of Damavand tokamak are modified. These circuits can control the vertical position of the plasma column inside the vacuum vessel. Since the vertical position of plasma is an unstable parameter, a direct closed loop system identification algorithm is performed. In the second stage, a nonlinear model is identified for plasma vertical position, based on the multilayer perceptron (MLP) neural network (NN) structure. Estimation of simulator parameters has been performed by back-propagation error algorithm using Levenberg-Marquardt gradient descent optimization technique. The model is verified through simulation of the whole closed loop system using both simulator and actual plant in similar conditions. As the final stage, a MLP neural network controller is designed for simulator model. In the last step, online training is performed to tune the controller parameters. Simulation results justify using of the NN controller for the actual plant. PMID:23464208
Identification and control of plasma vertical position using neural network in Damavand tokamak
NASA Astrophysics Data System (ADS)
Rasouli, H.; Rasouli, C.; Koohi, A.
2013-02-01
In this work, a nonlinear model is introduced to determine the vertical position of the plasma column in Damavand tokamak. Using this model as a simulator, a nonlinear neural network controller has been designed. In the first stage, the electronic drive and sensory circuits of Damavand tokamak are modified. These circuits can control the vertical position of the plasma column inside the vacuum vessel. Since the vertical position of plasma is an unstable parameter, a direct closed loop system identification algorithm is performed. In the second stage, a nonlinear model is identified for plasma vertical position, based on the multilayer perceptron (MLP) neural network (NN) structure. Estimation of simulator parameters has been performed by back-propagation error algorithm using Levenberg-Marquardt gradient descent optimization technique. The model is verified through simulation of the whole closed loop system using both simulator and actual plant in similar conditions. As the final stage, a MLP neural network controller is designed for simulator model. In the last step, online training is performed to tune the controller parameters. Simulation results justify using of the NN controller for the actual plant.
Tronko, Natalia; Vittot, Michel
2008-11-01
Localized hamiltonian control theory gives the possibility to reduce the radial chaotic transport of plasma test-particles into the Tokamak, by creating the Internal Transport Barrier(ITB). We prove that the control term is of quadratic order in the perturbation of Hamiltonian. We apply this method to a phenomenological model of electric potential in magnetized plasma.
Measurements of plasma composition in the TEXTOR tokamak by collective Thomson scattering
NASA Astrophysics Data System (ADS)
Stejner, M.; Korsholm, S. B.; Nielsen, S. K.; Salewski, M.; Bindslev, H.; Brezinsek, S.; Furtula, V.; Leipold, F.; Michelsen, P. K.; Meo, F.; Moseev, D.; Bürger, A.; Kantor, M.; de Baar, M.; TEXTOR Team
2012-01-01
We demonstrate the use of collective Thomson scattering (CTS) for spatially localized measurements of the isotopic composition of magnetically confined fusion plasmas. The experiments were conducted in the TEXTOR tokamak by scattering millimeter-wave probe radiation off plasma fluctuations with wave vector components nearly perpendicular to the magnetic field. Under such conditions the sensitivity of the CTS spectrum to plasma composition is enhanced by the spectral signatures of the ion cyclotron motion and of weakly damped ion Bernstein waves. Recent experiments on TEXTOR demonstrated the ability to resolve these signatures in the CTS spectrum as well as their sensitivity to the ion species mix in the plasma. This paper shows that the plasma composition can be inferred from the measurements through forward modeling of the CTS spectrum. We demonstrate that spectra measured in plasmas consisting of hydrogen, deuterium and 3He can be accurately reproduced by theory and yield inferred plasma compositions consistent with expectations. The potential to use CTS for measurements of plasma composition is of significant interest since CTS is well suited for reactor environments and since there is at present no established method to measure the fuel ion density ratio in the core of a burning fusion plasma.
Trapped Electron Mode Turbulence Driven Intrinsic Rotation in Tokamak Plasmas
Wang, W. X.; Hahm, T. S.; Ethier, S.; Zakharov, L. E.
2011-02-07
Recent progress from global gyrokinetic simulations in understanding the origin of intrinsic rotation in toroidal plasmas is reported with emphasis on electron thermal transport dominated regimes. The turbulence driven intrinsic torque associated with nonlinear residual stress generation by the fluctuation intensity and the intensity gradient in the presence of zonal flow shear induced asymmetry in the parallel wavenumber spectrum is shown to scale close to linearly with plasma gradients and the inverse of the plasma current. These results qualitatively reproduce empirical scalings of intrinsic rotation observed in various experiments. The origin of current scaling is found to be due to enhanced kll symmetry breaking induced by the increased radial variation of the safety factor as the current decreases. The physics origin for the linear dependence of intrinsic torque on pressure gradient is that both turbulence intensity and the zonal flow shear, which are two key ingredients for driving residual stress, increase with the strength of turbulence drive, which is R0/LTe and R0/Lne for the trapped electron mode. __________________________________________________
Linear multispecies gyrokinetic flux tube benchmarks in shaped tokamak plasmas
NASA Astrophysics Data System (ADS)
Merlo, G.; Sauter, O.; Brunner, S.; Burckel, A.; Camenen, Y.; Casson, F. J.; Dorland, W.; Fable, E.; Görler, T.; Jenko, F.; Peeters, A. G.; Told, D.; Villard, L.
2016-03-01
Verification is the fundamental step that any turbulence simulation code has to be submitted in order to assess the proper implementation of the underlying equations. We have carried out a cross comparison of three flux tube gyrokinetic codes, GENE [F. Jenko et al., Phys. Plasmas 7, 1904 (2000)], GKW [A. G. Peeters et al., Comput. Phys. Commun. 180, 2650 (2009)], and GS2 [W. Dorland et al., Phys. Rev. Lett. 85, 5579 (2000)], focusing our attention on the effect of realistic geometries described by a series of MHD equilibria with increasing shaping complexity. To simplify the effort, the benchmark has been limited to the electrostatic collisionless linear behaviour of the system. A fully gyrokinetic model has been used to describe the dynamics of both ions and electrons. Several tests have been carried out looking at linear stability at ion and electron scales, where for the assumed profiles Ion Temperature Gradient (ITG)/Trapped Electron Modes and Electron Temperature Gradient modes are unstable. The capability of the codes to handle a non-zero ballooning angle has been successfully benchmarked in the ITG regime. Finally, the standard Rosenbluth-Hinton test has been successfully carried out looking at the effect of shaping on Zonal Flows (ZFs) and Geodesic Acoustic Modes (GAMs). Inter-code comparison as well as validation of simulation results against analytical estimates has been accomplished. All the performed tests confirm that plasma elongation strongly stabilizes plasma instabilities as well as leads to a strong increase in ZF residual and GAM damping.
Superconducting magnet system for the TPX Tokamak
Hassenzahl, W.V.; Chaplin, M.R.; Heim, J.R.
1993-09-15
The Tokamak Physics Experiment (TPX) will be the first Tokamak using superconducting magnets for both the poloidal and toroidal field. It is designed for advanced Tokamak physics experiments in steady-state and long-pulse operation. The TPX superconducting magnets use an advanced cable-in-conduit conductor (CICC) design similar to that developed in support of the International Thermonuclear Experimental Reactor (ITER). The toroidal field magnets provide 4.0 T at 2.25 m with a stored energy of 1.05 GJ. The poloidal field magnets provide 18.0 V-s to ohmically start and control long burns of a 2.0 MA plasma.
Scattering of radio frequency waves by cylindrical density filaments in tokamak plasmas
NASA Astrophysics Data System (ADS)
Ram, Abhay K.; Hizanidis, Kyriakos
2016-02-01
In tokamak fusion plasmas, coherent fluctuations in the form of blobs or filaments are routinely observed in the scrape-off layer. Radio frequency (RF) electromagnetic waves, excited by antenna structures placed near the wall of a tokamak, have to propagate through the scrape-off layer before reaching the core of the plasma. While the effect of fluctuations on the properties of RF waves has not been quantified experimentally, it is of interest to carry out a theoretical study to determine if fluctuations can affect the propagation characteristics of RF waves. Usually, the difference between the plasma density inside the filament and the background plasma density is sizable, the ratio of the density difference to the background density being of order one. Generally, this precludes the use of geometrical optics in determining the effect of fluctuations, since the relevant ratio has to be much less than one, typically, of the order of 10% or less. In this paper, a full-wave, analytical model is developed for the scattering of a RF plane wave by a cylindrical plasma filament. It is assumed that the plasma inside and outside the filament is cold and uniform and that the major axis of the filament is aligned along the toroidal magnetic field. The ratio of the density inside the filament to the density of the background plasma is not restricted. The theoretical framework applies to the scattering of any cold plasma wave. In order to satisfy the boundary conditions at the interface between the filament and the background plasma, the electromagnetic fields inside and outside the filament need to have the same k∥ , the wave vector parallel to the ambient magnetic field, as the incident plane wave. Consequently, in contrast to the scattering of a RF wave by a spherical blob [Ram et al., Phys. Plasmas 20, 056110-1-056110-10 (2013)], the scattering by a field-aligned filament does not broaden the k∥ spectrum. However, the filament induces side-scattering leading to surface
NASA Astrophysics Data System (ADS)
Murari, A.; Peluso, E.; Gelfusa, M.; Lupelli, I.; Gaudio, P.
2015-07-01
The extrapolation of the energy confinement time to the next generation of devices has been investigated both theoretically and experimentally for several decades in the tokamak community. Various scaling expressions have been proposed using dimensional and dimensionless quantities. They are all based on the assumption that the scalings are in power law form. In this paper, an innovative methodology is proposed to extract the scaling expressions for the energy confinement time in tokamaks directly from experimental databases, without any previous assumption about the mathematical form of the scalings. The approach to obtain the scaling expressions is based on genetic programming and symbolic regression. These techniques have been applied to the ITPA database of H-mode discharges and the results have been validated with a series of established statistical tools. The soundest results, using dimensional variables, are not in the form of power laws but contain a multiplicative saturation term. Also the scalings, expressed in terms of the traditional dimensionless quantities, are not in power law form and contain additive saturation terms. The extrapolation to ITER of both dimensional and dimensionless quantities indicate that the saturation effects are quite significant and could imply a non-negligible reduction in the confinement time to be expected in the next generation of devices. The results obtained with the proposed techniques therefore motivate a systematic revisiting of the scaling expressions for plasma confinement in tokamaks.
Hypersonic drift-tearing magnetic islands in tokamak plasmas
Fitzpatrick, R.; Waelbroeck, F. L.
2007-12-15
A two-fluid theory of long wavelength, hypersonic, drift-tearing magnetic islands in low-collisionality, low-{beta} plasmas possessing relatively weak magnetic shear is developed. The model assumes both slab geometry and cold ions, and neglects electron temperature and equilibrium current gradient effects. The problem is solved in three asymptotically matched regions. The 'inner region' contains the island. However, the island emits electrostatic drift-acoustic waves that propagate into the surrounding 'intermediate region', where they are absorbed by the plasma. Since the waves carry momentum, the inner region exerts a net force on the intermediate region, and vice versa, giving rise to strong velocity shear in the region immediately surrounding the island. The intermediate region is matched to the surrounding 'outer region', in which ideal magnetohydrodynamic holds. Isolated hypersonic islands propagate with a velocity that lies between those of the unperturbed local ion and electron fluids, but is much closer to the latter. The ion polarization current is stabilizing, and increases with increasing island width. Finally, the hypersonic branch of isolated island solutions ceases to exist above a certain critical island width. Hypersonic islands whose widths exceed the critical width are hypothesized to bifurcate to the so-called 'sonic' solution branch.
Modification of plasma flows with gas puff in the scrape-off layer of ADITYA tokamak
Sangwan, Deepak; Jha, Ratneshwar; Brotankova, Jana; Gopalkrishna, M. V.
2013-06-15
The parallel Mach numbers are measured at three locations in the scrape-off layer (SOL) plasma of ADITYA tokamak by using Mach probes. The flow pattern is constructed from these measurements and the modification of flow pattern is observed by introducing a small puff of working gas. In the normal discharge, there is an indication of shell structure in the SOL plasma flows, which is removed during the gas puff. The plasma parameters, particle flux and Reynolds stress are also measured in the normal discharge and in the discharge with gas puff. It is observed that Reynolds stress and Mach number are coupled in the near SOL region and decoupled in the far SOL region. The coupling in the near SOL region gets washed away during the gas puff.
Effects of high Z probe on plasma behavior in HT-6M tokamak
NASA Astrophysics Data System (ADS)
Li, J.; Gong, X.; Luo, L.; Yin, F. X.; Noda, N.; Wan, B.; Xu, W.; Gao, X.; Yin, F.; Jiang, J. G.; Wu, Z.; Zhao, J. Y.; Wu, M.; Liu, S.; Han, Y.
1997-02-01
Molybdenum and tungsten probes have been tested in HT-6M tokamak under various discharge conditions aiming to find out the conditions in which high Z PFC can be used without serious degradation of core plasma performance. In normal OH discharges, the degradation of core plasma performance was found only when the probe was inserted beyond 3.0 cm inside the last closed flux surface (LCFS). The plasma performance did not change with positive biasing to the probe, whereas central Te degraded during negative biasing of -100 V. The insertion of the Mo probe to 1.5 cm inside the LCFS made a change in the threshold power of the L-H transition in EOH discharges. These results suggest a certain operation range of the H-mode in the EOH discharge with the Mo probe in HT-6M.
Damping of electron cyclotron waves in dense plasmas of a compact ignition tokamak
Mazzucato, E.; Fidone, I.; Granata, G.
1987-06-01
Absorption of electromagnetic waves by hot and dense plasmas is investigated in the electron cyclotron range of frequency. It is shown that the strong reduction of the damping of the extraordinary mode, caused by finite Larmor radius effects on waves propagating perpendicularly to the magnetic field, becomes insignificant at large values of the parallel component of the refractive index. With an appropriate form of the relativistic dispersion relation which includes high order Larmor radius terms, heating of dense plasmas in a Compact Ignition Tokamak is investigated. It is shown that by using the extraordinary mode with oblique propagation and frequency of 190 GHz it is possible to bring to thermonuclear ignition a dense ohmic plasma with a toroidal magnetic field of 105 kG and a central density of 1 x 10/sup 15/ cm/sup -3/. 11 refs., 11 figs.
Charge-exchange recombination spectroscopy of the plasma ion temperature at the T-10 tokamak
Krupin, V. A.; Tugarinov, S. N.; Barsukov, A. G.; Dnestrovskij, A. Yu.; Klyuchnikov, L. A.; Korobov, K. V.; Krasnyanskii, S. A.; Naumenko, N. N.; Nemets, A. R.; Sushkov, A. V.; Tilinin, G. N.
2013-08-15
Charge-exchange recombination spectroscopy (CXRS) based on a diagnostic neutral beam has been developed at the T-10 tokamak. The diagnostics allows one to measure the ion temperature profile in the cross section of the plasma column. In T-10 experiments, the measurement technique was adjusted and the elements of the CXRS diagnostics for ITER were tested. The used spectroscopic equipment makes it possible to reliably determine the ion temperature from the Doppler broadening of impurity lines (helium, carbon), as well as of the spectral lines of the working gas. The profiles of the plasma ion temperature in deuterium and helium discharges were measured at different plasma currents and densities, including with the use of active Doppler measurements of lines of different elements. The validity and reliability of ion temperature measurements performed by means of the developed CXRS diagnostics are analyzed.
A new asymmetric Abel-inversion method for plasma interferometry in tokamaks
Park, H.K.
1989-02-01
In order to get precise local electron density information from chordal interferometric measurement of a tokamak plasma, a self- consistent and reliable inversion method is necessary. In this paper, a new asymmetric Abel-inversion method is introduced. This method includes flexible boundary conditions, application to a non-circular geometry, and estimation of the plasma in the scrape-off layer. The advantages of this method are demonstrated by comparison with other methods. This new inversion method is applied to a parametric study which includes dependence on the Shafranov shift and elongation of the profile. The inverted results are integrated along different views and compared with other density measurements. This new method can also be applied to plasma spectroscopy. 6 refs., 6 figs.
Calculated radiative power losses from mid- and high-Z impurities in Tokamak plasmas
NASA Astrophysics Data System (ADS)
Fournier, Kevin B.; May, M. J.; Pacella, D.; Gregory, B. C.; Rice, J. E.; Terry, J. L.; Finkenthal, M.; Goldstein, W. H.
1998-09-01
This paper summarizes recent calculations of the radiative cooling coefficient for molybdenum (Z=42), krypton (Z=36) and argon (Z=18). The radiative processes considered are collisional-radiative line emission, dielectronic recombination line emission, and radiative recombination and bremsstrahlung continuum emission. Collisional-radiative line emission dominates the power loss channels for a given impurity at all but the highest plasma electron temperatures. The atomic data for the line emission are computed ab initio with the HULLAC atomic physics suite of codes. Relativistic, ab initio atomic physics data are used to compute ionization and recombination rate coefficients; the resulting charge state distribution and recombination rates are used to estimate the radiative power from recombination processes. The calculations in the present work are benchmarked against absolute measurements of ion brightness profiles in the Frascati Tokamak Upgrade plasma. Integrated measurements from tokamak plasmas such as bolometry are then simulated. The atomic physics data used to predict the emissivity of individual ions is validated; the calculated cooling coefficients agree well with bolometric measurements.
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe.
Walkden, N R; Adamek, J; Allan, S; Dudson, B D; Elmore, S; Fishpool, G; Harrison, J; Kirk, A; Komm, M
2015-02-01
The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ∼1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the ER measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak. PMID:25725845
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe
Walkden, N. R.; Adamek, J.; Komm, M.; Allan, S.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Dudson, B. D.
2015-02-15
The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ∼1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the E{sub R} measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak.
Park, Hyeon, K.; Sabbagh, S.A.
1996-05-01
The hypothesis that the heating beam fueling profile shape connects the edge condition and improved core confinement and fusion reactivity is extensively studied on TFTR and applied to other tokamaks. The derived absolute scalings based on beam fueling profile shape for the stored energy and neutron yield can be applied to the deuterium discharges at different major radii in TFTR. These include Supershot, High poloidal beta, L-mode, and discharges with a reversed shear (RS) magnetic configuration. These scalings are also applied to deuterium-tritium discharges. The role of plasma parameters, such as plasma current, Isdo2(p), edge safety factor, qsdo5(a), and toroidal field, Bsdo2(T), in the performance and stability of the discharges is explicitly studied. Based on practical and externally controllable plasma parameters, the limitation and optimization of fusion power production of the present TFTR is investigated and a path for a discharge condition with fusion power gain, Q > 1 is suggested based on this study. Similar physics interpretation is provided for beam heated discharges on other major tokamaks.
Shaping of the plasma column in a small aspect ratio tokamak
NASA Astrophysics Data System (ADS)
Herrera, Julio; Arroyo, Ismael; Chavez, Esteban
2015-11-01
This is a follow-up to the work presented in last year's meeting, on the conceptual design of a small aspect ratio tokamak of variable configuration. The base parameters for this device would be similar to those in the START tokamak. The shaping of the plasma column is known to have important effects in the plasma performance, including the value of β, bootstrap currents, and intrinsic rotation. The main feature being explored here is the inclusion of independent control coils in the inboard and outboard sides; six in the first case, and up to seven in the latter. By varying the strength in their currents it is possible to achieve a wide variety of shapes: elliptical, conventional D-shape, inverse D-shape, and Bean-shape. As the control coils are activated, the strength of the toroidal magnetic field needs to he weakened, in order to keep reasonable values of the safety factor q . The study presented here is made by means of the 3D-MAPTOR code, which produces the Poincaré maps of the magnetic field lines, given the currents. For this purpose, a seed plasma current must be provided. All studies presented here assume equatorial symmetry, due to limitations in the code.
Effect of Ion Cyclotron Heating on Fast Ion Transport and Plasma Rotation in Tokamaks
NASA Astrophysics Data System (ADS)
Chan, V. S.; Omelchenko, Y. A.; Chiu, S. C.
2000-10-01
Minority ion cyclotron heating can produce energetic ions with banana orbits which are finite compared with the minor radius of a tokamak. The radial transport of the fast ions in the presence of Coulomb collisions results in a radial current and a corresponding JxB torque density on the bulk plasma. Collisions with the bulk ions provides an additional frictional torque that adds to or opposes the magnetic torque. This study clarifies the various mechanisms which can contribute to the torque components including collision-induced finite orbit particle diffusion, wave-induced asymmetry in canonical momentum when doppler resonance is accounted for, and orbit asymmetry created by magnetic geometry. Ion dynamics are calculated with a Monte-Carlo code in which wave-induced energy diffusion is accounted for by a quasilinear operator. The code follows particle drift trajectories in a tokamak geometry under the influence of RF fields and collisions with the background plasma. Questions on the direction of plasma rotation under different conditions and validity of the Green's function approach in modeling RF-induced rotation will be addressed.
Profile measurements in the plasma edge of mega amp spherical tokamak using a ball pen probe
NASA Astrophysics Data System (ADS)
Walkden, N. R.; Adamek, J.; Allan, S.; Dudson, B. D.; Elmore, S.; Fishpool, G.; Harrison, J.; Kirk, A.; Komm, M.
2015-02-01
The ball pen probe (BPP) technique is used successfully to make profile measurements of plasma potential, electron temperature, and radial electric field on the Mega Amp Spherical Tokamak. The potential profile measured by the BPP is shown to significantly differ from the floating potential both in polarity and profile shape. By combining the BPP potential and the floating potential, the electron temperature can be measured, which is compared with the Thomson scattering (TS) diagnostic. Excellent agreement between the two diagnostics is obtained when secondary electron emission is accounted for in the floating potential. From the BPP profile, an estimate of the radial electric field is extracted which is shown to be of the order ˜1 kV/m and increases with plasma current. Corrections to the BPP measurement, constrained by the TS comparison, introduce uncertainty into the ER measurements. The uncertainty is most significant in the electric field well inside the separatrix. The electric field is used to estimate toroidal and poloidal rotation velocities from E × B motion. This paper further demonstrates the ability of the ball pen probe to make valuable and important measurements in the boundary plasma of a tokamak.
Plasma edge transport with magnetic islands—a comparison between tokamak and reversed-field pinch
NASA Astrophysics Data System (ADS)
Ciaccio, G.; Schmitz, O.; Abdullaev, S. S.; Frerichs, H.; Agostini, M.; Scarin, P.; Spizzo, G.; Vianello, N.; White, R. B.
2014-06-01
In the reversed-field pinch (RFP) edge, measured transport and flows are strongly influenced by magnetic islands (Vianello 2013 Nucl. Fusion 53 073025). In fact, these islands determine a differential radial diffusion of electrons and ions which, interacting with the wall, give rise to a characteristic edge ambipolar potential. Such island structures also arise in tokamak plasmas, when resonant magnetic perturbations (RMPs) are applied for control of edge-localized modes. They impose a characteristic modulation to edge electron density and temperature fields, in close correlation with the local magnetic vacuum topology (Schmitz 2012 Nucl. Fusion 52 054001). In order to develop a generic picture of particle transport with magnetic islands located in the plasma edge between RFPs and tokamaks with RMP, test-particle transport simulations are done on TEXTOR with the same tool used in RFX-mod, namely, the guiding-centre code ORBIT (White and Chance 1984 Phys. Fluids 27 2455-67). A typical TEXTOR discharge in the (m, n) = (12, 4) configuration is reconstructed and analysed with ORBIT. We use Poincaré and connection length analysis of electrons and ion orbits to analyse the magnetic structure taking into account the different gyro-orbits of both constituents. Density distributions of test ions and electrons are calculated and used to obtain an initial estimate of the plasma potential and radial electric field around the island.
FIRST MEASUREMENT OF PRESSURE GRADIENT-DRIVEN CURRENTS IN TOKAMAK EDGE PLASMAS
THOMAS DM; LEONARD AW; LAO LL; OSBORNE TH; MUELLER HW; FINKENTHAL DK
2003-11-01
Localized currents driven by pressure gradients play a pivotal role in the magnetohydrodynamic stability of toroidal plasma confinement devices. We have measured the currents generated in the edge of L- (low) and H- (high confinement) mode discharges on the DIII-D tokamak, utilizing the Zeeman effect in an injected lithium beam to obtain high resolution profiles of the poloidal magnetic field. We find current densities in excess of 1 MA/m{sup 2} in a 1 to 2 cm region near the peak of the edge pressure gradient. These values are sufficient to challenge edge stability theories based on specific current formation models.
Effects of magnetic shear on toroidal rotation in tokamak plasmas with lower hybrid current drive.
Rice, J E; Podpaly, Y A; Reinke, M L; Mumgaard, R; Scott, S D; Shiraiwa, S; Wallace, G M; Chouli, B; Fenzi-Bonizec, C; Nave, M F F; Diamond, P H; Gao, C; Granetz, R S; Hughes, J W; Parker, R R; Bonoli, P T; Delgado-Aparicio, L; Eriksson, L-G; Giroud, C; Greenwald, M J; Hubbard, A E; Hutchinson, I H; Irby, J H; Kirov, K; Mailloux, J; Marmar, E S; Wolfe, S M
2013-09-20
Application of lower hybrid (LH) current drive in tokamak plasmas can induce both co- and countercurrent directed changes in toroidal rotation, depending on the core q profile. For discharges with q(0) <1, rotation increments in the countercurrent direction are observed. If the LH-driven current is sufficient to suppress sawteeth and increase q(0) above unity, the core toroidal rotation change is in the cocurrent direction. This change in sign of the rotation increment is consistent with a change in sign of the residual stress (the divergence of which constitutes an intrinsic torque that drives the flow) through its dependence on magnetic shear. PMID:24093268
Electron cyclotron emission in a rf current-driven tokamak plasma
Giruzzi, G.; Fidone, I.; Granata, G.; Meyer, R.L.
1984-07-01
The general properties of electron cyclotron radiation emitted by the current-carrying superthermal tail in tokamak plasmas are investigated. Two situations are considered, namely, the case of an extended tail as it occurs in present-day experiments and that in which the velocity range of the tail is much less than its mean speed. For the former case, we show that most of the emitted radiation obeys Kirchhoff's law. In the latter case, which is suited for reactor-like parameters, a simple relation is obtained between the tail momentum distribution and the radiation temperature.
Kinetic theory and simulation of multi-species plasmas in tokamaks excited with ICRF microwaves
Kerbel, G.D.; McCoy, M.G.
1984-12-21
This paper presents a description of a bounce-averaged Fokker-Planck quasilinear model for the kinetic description of tokamak plasmas. The non-linear collision and quasilinear resonant diffusion operators are represented in a form conducive to numerical solution with specific attention to the treatment of the boundary layer separating trapped and passing orbit regions of velocity space. The numerical techniques employed are detailed in so far as they constitute significant departure from those used in the conventional uniform magnetic field case. Examples are given to illustrate the combined effects of collisional and resonant diffusion.
Geodesic acoustic modes in tokamak plasmas with a radial equilibrium electric field
Zhou, Deng
2015-09-15
The dispersion relation of geodesic acoustic modes in the tokamak plasma with an equilibrium radial electric field is derived and analyzed. Multiple branches of eigenmodes have been found, similar to the result given by the fluid model with a poloidal mass flow. Frequencies and damping rates of both the geodesic acoustic mode and the sound wave increase with respect to the strength of radial electric field, while the frequency and the damping rate of the lower frequency branch slightly decrease. Possible connection to the experimental observation is discussed.
The effect of ion drifts on the properties of the tokamak scrape-off plasma
Petravic, M.; Kuo-Petravic, G.
1988-09-01
A plasma fluid model which takes into account ion drifts has been constructed and applied to the scrape-off layer of a tokamak with a poloidal divertor. This model predicts near-sonic toroidal velocities and large poloidal flows in most of the scrapeoff together with steep gradients in the pressure and electrostatic potential along the magnetic field near the X-point, contrary to the predictions of the standard model. The potential step at X-point should reduce parallel heat transport and could act as an H-mode trigger. 12 refs., 4 figs.
On the non-stiffness of edge transport in L-mode tokamak plasmas
Sauter, O.; Brunner, S.; Kim, D.; Merlo, G.; Behn, R.; Coda, S.; Duval, B. P.; Federspiel, L.; Goodman, T. P.; Karpushov, A.; Merle, A.; Team, TCV; Camenen, Y.
2014-05-15
Transport analyses using first-principle turbulence codes and 11/2 -D transport codes usually study radial transport properties between the tokamak plasma magnetic axis and a normalized minor radius around 0.8. In this region, heat transport shows significantly stiff properties resulting in temperature scalelength values (R∕L{sub T}) that are relatively independent of the level of the radial heat flux. We have studied experimentally in the tokamak à configuration variable [F. Hofmann et al., Plasma Phys. Controlled Fusion 36, B277 (1994)] the radial electron transport properties of the edge region, close to the last closed flux surface, namely, between ρ{sub V}=√(V/V{sub edge})=0.8 to 1. It is shown that electron transport is not stiff in this region and high R∕L{sub Te} values (∼20–40) can be attained even for L-mode confinement. We can define a “pedestal” location, already in L-mode regimes, where the transport characteristics change from constant logarithmic gradient, inside ρ{sub V} = 0.8, to constant gradient between 0.8 and 1.0. In particular, we demonstrate, with well resolved T{sub e} and n{sub e} profiles, that the confinement improvement with plasma current I{sub p}, with or without auxiliary heating, is due to this non-stiff edge region. This new result is used to explain the significant confinement improvement observed with negative triangularity, which could not be explained by theory to date. Preliminary local gyrokinetic simulations are now consistent with an edge, less stiff, region that is more sensitive to triangularity than further inside. We also show that increasing the electron cyclotron heating power increases the edge temperature inverse scalelength, in contrast to the value in the main plasma region. The dependence of confinement on density in ohmic plasmas is also studied and brings new insight in the understanding of the transition between linear and saturated confinement regimes, as well as of the density limit and
Conceptual design of a fast-ion D-alpha diagnostic on experimental advanced superconducting tokamak
Huang, J. Wan, B.; Hu, L.; Hu, C.; Heidbrink, W. W.; Zhu, Y.; Hellermann, M. G. von; Gao, W.; Wu, C.; Li, Y.; Fu, J.; Lyu, B.; Yu, Y.; Ye, M.; Shi, Y.
2014-11-15
To investigate the fast ion behavior, a fast ion D-alpha (FIDA) diagnostic system has been planned and is presently under development on Experimental Advanced Superconducting Tokamak. The greatest challenges for the design of a FIDA diagnostic are its extremely low intensity levels, which are usually significantly below the continuum radiation level and several orders of magnitude below the bulk-ion thermal charge-exchange feature. Moreover, an overlaying Motional Stark Effect (MSE) feature in exactly the same wavelength range can interfere. The simulation of spectra code is used here to guide the design and evaluate the diagnostic performance. The details for the parameters of design and hardware are presented.
Topology of tokamak plasma equilibria with toroidal current reversal
Rodrigues, Paulo; Bizarro, Joao P. S.
2012-01-15
Some general principles about scalar functions with critical points are used to rigorously ascertain that magnetic equilibria with both toroidal current reversal and nested magnetic surfaces are atypical solutions and highly unstable to arbitrary perturbations of boundary conditions and other parameters. The cause for such is shown to lie in the condition of nested magnetic surfaces and not in the possibility of current reversal and consequent vanishing of the poloidal field inside the plasma. Rather than supporting the claim that instability against experimentally driven perturbations forbids configurations with toroidal current reversal, it is argued that these can be attained if an axisymmetric island system is allowed for in order to break the condition of nested magnetic surfaces. A number of results previously reported in the literature are discussed and reinterpreted under the proposed framework, providing some physical insight on the nature of equilibria with toroidal current reversal.
Runaway electron dynamics in tokamak plasmas with high impurity content
Martín-Solís, J. R.; Loarte, A.; Lehnen, M.
2015-09-15
The dynamics of high energy runaway electrons is analyzed for plasmas with high impurity content. It is shown that modified collision terms are required in order to account for the collisions of the relativistic runaway electrons with partially stripped impurity ions, including the effect of the collisions with free and bound electrons, as well as the scattering by the full nuclear and the electron-shielded ion charge. The effect of the impurities on the avalanche runaway growth rate is discussed. The results are applied, for illustration, to the interpretation of the runaway electron behavior during disruptions, where large amounts of impurities are expected, particularly during disruption mitigation by massive gas injection. The consequences for the electron synchrotron radiation losses and the resulting runaway electron dynamics are also analyzed.
Bifurcations of axisymmetric plasma equilibrium in a tokamak
NASA Astrophysics Data System (ADS)
Skovoroda, A. A.
2016-05-01
Bifurcation of solutions to the Grad-Shafranov-type equation for helically symmetric plasma near the threshold for tearing instability are analyzed. Quadratic and cubic nonlinearities were added to the linear dependence of the current density on the helical flux. Depending on the character of nonlinearity, two types of bifurcation can be observed, the "small" and the "large" ones. The small bifurcation is typical of cubic nonlinearity and reveals itself in the growth of the magnetic island from zero as the profile parameter increases above the instability threshold. The large bifurcation is typical of quadratic nonlinearity and causes jumplike formation of a large-scale magnetic island upon exceeding the instability threshold. As the profile parameter decreases below the instability threshold, the large-scale island continues to persist for some time (the hysteresis effect) and then suddenly disappears.
Runaway electron dynamics in tokamak plasmas with high impurity content
NASA Astrophysics Data System (ADS)
Martín-Solís, J. R.; Loarte, A.; Lehnen, M.
2015-09-01
The dynamics of high energy runaway electrons is analyzed for plasmas with high impurity content. It is shown that modified collision terms are required in order to account for the collisions of the relativistic runaway electrons with partially stripped impurity ions, including the effect of the collisions with free and bound electrons, as well as the scattering by the full nuclear and the electron-shielded ion charge. The effect of the impurities on the avalanche runaway growth rate is discussed. The results are applied, for illustration, to the interpretation of the runaway electron behavior during disruptions, where large amounts of impurities are expected, particularly during disruption mitigation by massive gas injection. The consequences for the electron synchrotron radiation losses and the resulting runaway electron dynamics are also analyzed.
Tokamak Physics Experiment (TPX) power supply design and development
Neumeyer, C.; Bronner, G.; Lu, E.; Ramakrishnan, S.
1995-04-01
The Tokamak Physics Experiment (TPX) is an advanced tokamak project aimed at the production of quasi-steady state plasmas with advanced shape, heating, and particle control. TPX is to be built at the Princeton Plasma Physics Laboratory (PPPL) using many of the facilities from the Tokamak Fusion Test Reactor (TFTR). TPX will be the first tokamak to utilize superconducting (SC) magnets in both the toroidal field (TF) and poloidal field (PF) systems. This new feature requires a departure from the traditional tokamak power supply schemes. This paper describes the plan for the adaptation of the PPPL/FTR power system facilities to supply TPX. Five major areas are addressed, namely the AC power system, the TF, PF and Fast Plasma Position Control (FPPC) power supplies, and quench protection for the TF and PF systems. Special emphasis is placed on the development of new power supply and protection schemes.
A penalization technique to model plasma facing components in a tokamak with temperature variations
NASA Astrophysics Data System (ADS)
Paredes, A.; Bufferand, H.; Ciraolo, G.; Schwander, F.; Serre, E.; Ghendrih, P.; Tamain, P.
2014-10-01
To properly address turbulent transport in the edge plasma region of a tokamak, it is mandatory to describe the particle and heat outflow on wall components, using an accurate representation of the wall geometry. This is challenging for many plasma transport codes, which use a structured mesh with one coordinate aligned with magnetic surfaces. We propose here a penalization technique that allows modeling of particle and heat transport using such structured mesh, while also accounting for geometrically complex plasma-facing components. Solid obstacles are considered as particle and momentum sinks whereas ionic and electronic temperature gradients are imposed on both sides of the obstacles along the magnetic field direction using delta functions (Dirac). Solutions exhibit plasma velocities (M=1) and temperatures fluxes at the plasma-wall boundaries that match with boundary conditions usually implemented in fluid codes. Grid convergence and error estimates are found to be in agreement with theoretical results obtained for neutral fluid conservation equations. The capability of the penalization technique is illustrated by introducing the non-collisional plasma region expected by the kinetic theory in the immediate vicinity of the interface, that is impossible when considering fluid boundary conditions. Axisymmetric numerical simulations show the efficiency of the method to investigate the large-scale transport at the plasma edge including the separatrix and in realistic complex geometries while keeping a simple structured grid.
Spectra of germanium and selenium in the 50-350 A region from the PLT tokamak plasma
Stratton, B.C.; Hodge, W.L.; Moos, H.W.; Schwob, J.L.; Suckewer, S.; Finkenthal, M.; Cohen, S.
1983-03-01
Spectra of germanium and selenium injected into the PLT tokamak plasma were observed in the 50 to 350 A region for GeXIV-XXV (KI to OI-like) and SeXVI-XXIV (KI to NaI-like). A number of 3p/sup k/-3p/sup k-1/3d transitions predicted by isoelectronic sequence extrapolation have been identified. Also, previously identified lines from ions in the AlI to OI-like and KI-like isoelectronic sequences have been observed in the tokamak plasma.
Isotope mass and charge effects in tokamak plasmas
Pusztai, I.; Candy, J.; Gohil, P.
2011-12-15
The effect of primary ion species of differing charge and mass - specifically, deuterium, hydrogen, and helium - on instabilities and transport is studied in DIII-D plasmas through gyrokinetic simulations with gyro [J. Candy and E. Belli, General Atomics Technical Report No. GA-A26818, 2010]. In linear simulations under imposed similarity of the profiles, there is an isomorphism between the linear growth rates of hydrogen isotopes, but the growth rates are higher for Z > 1 main ions due to the appearance of the charge in the Poisson equation. On ion scales the most significant effect of the different electron-to-ion mass ratio appears through collisions stabilizing trapped electron modes. In nonlinear simulations, significant favorable deviations from pure gyro-Bohm scaling are found due to electron-to-ion mass ratio effects and collisions. The presence of any non-trace impurity species cannot be neglected in a comprehensive simulation of the transport; including carbon impurity in the simulations caused a dramatic reduction of energy fluxes. The transport in the analyzed deuterium and helium discharges could be well reproduced in gyrokinetic and gyrofluid simulations while the significant hydrogen discrepancy is the subject of ongoing investigation.
Isotope mass and charge effects in tokamak plasmas
NASA Astrophysics Data System (ADS)
Pusztai, I.; Candy, J.; Gohil, P.
2011-12-01
The effect of primary ion species of differing charge and mass—specifically, deuterium, hydrogen, and helium—on instabilities and transport is studied in DIII-D plasmas through gyrokinetic simulations with gyro [J. Candy and E. Belli, General Atomics Technical Report No. GA-A26818, 2010]. In linear simulations under imposed similarity of the profiles, there is an isomorphism between the linear growth rates of hydrogen isotopes, but the growth rates are higher for Z > 1 main ions due to the appearance of the charge in the Poisson equation. On ion scales the most significant effect of the different electron-to-ion mass ratio appears through collisions stabilizing trapped electron modes. In nonlinear simulations, significant favorable deviations from pure gyro-Bohm scaling are found due to electron-to-ion mass ratio effects and collisions. The presence of any non-trace impurity species cannot be neglected in a comprehensive simulation of the transport; including carbon impurity in the simulations caused a dramatic reduction of energy fluxes. The transport in the analyzed deuterium and helium discharges could be well reproduced in gyrokinetic and gyrofluid simulations while the significant hydrogen discrepancy is the subject of ongoing investigation.
Studies of Feedback Stabilization of Axisymmetric Modes in Deformable Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Ward, David John
A new linear MHD stability code, NOVA-W, is described and applied to the study of the feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA^1 that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The code has been tested for the case of passive stabilization against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. The utility and performance of the NOVA-W code are demonstrated for calculations of varying configurations of passive conductors. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops which provide the measurements of vertical displacement. The results compare well to those of earlier calculations using a less efficient nonlinear code. The NOVA-W code is used to examine the effects of plasma deformability on feedback stabilization. It is seen that plasmas with shaped cross sections have unstable motion different from a rigid shift. Plasma equilibria with large triangularity show particularly significant deviations from a uniform rigid shift. Furthermore, the placement of passive conductors is shown to modify the non-rigid components of the motion in a way that reduces the stabilizing effects of these conductors. The eigenfunction is also modified under the effects of active feedback. This deformation is seen to depend strongly on the
M. Murakami; H.E. St.John; T.A. Casper; M.S. Chu; J.C. DeBoo; C.M. Greenfield; J.E. Kinsey; L.L. Lao; R.J. La Haye; Y.R. Lin-Liu; T.C. Luce; P.A. Politzer; B.W. Rice; G.M. Staebler; T.S. Taylor; M.R. Wade
1999-12-01
The status of modeling work focused on developing the advanced tokamak scenarios in DIII-D is discussed. The objectives of the work are two-fold: (1) to develop AT scenarios with ECCD using time-dependent transport simulations, coupled with heating and current drive models, consistent with MHD equilibrium and stability; and (2) to use time-dependent simulations to help plan experiments and to understand the key physics involved. Time-dependent simulations based on transport coefficients derived from experimentally achieved target discharges are used to perform AT scenario modeling. The modeling indicates off-axis ECCD with approximately 3 MW absorbed power can maintain high-performance discharges with q{sub min} > 1 for 5 to 10 s. The resultant equilibria are calculated to be stable to n = 1 pressure driven modes. The plasma is well into the second stability regime for high-n ballooning modes over a large part of the plasma volume. The role of continuous localized ECCD is studied for stabilizing m/n = 2/1 tearing modes. The progress towards validating current drive and transport models, consistent with experimental results, and developing self-consistent, integrated high performance AT scenarios is discussed.
Plasma Heating: An Advanced Technology
NASA Technical Reports Server (NTRS)
1994-01-01
The Mercury and Apollo spacecraft shields were designed to protect astronauts from high friction temperatures (well over 2,000 degrees Fahrenheit) when re-entering the Earth's atmosphere. It was necessary to test and verify the heat shield materials on Earth before space flight. After exhaustive research and testing, NASA decided to use plasma heating as a heat source. This technique involves passing a strong electric current through a rarefied gas to create a plasma (ionized gas) that produces an intensely hot flame. Although NASA did not invent the concept, its work expanded the market for commercial plasma heating systems. One company, Plasma Technology Corporation (PTC), was founded by a member of the team that developed the Re-entry Heating Simulator at Ames Research Center (ARC). Dr. Camacho, President of PTC, believes the technology has significant environmental applications. These include toxic waste disposal, hydrocarbon, decomposition, medical waste disposal, asbestos waste destruction, and chemical and radioactive waste disposal.
Recent Advances in Plasma Propulsion for Spacecraft
NASA Astrophysics Data System (ADS)
Choueiri, E. Y.
1998-11-01
Three decades of research on plasma propulsion for spacecraft have led to a level of maturity that has recently ushered in the era of application. Over the past few years, due to their ability to insure great mass savings over chemical propulsion, plasma propulsion devices (thrusters) have been used (or selected for use) on spacecraft for missions ranging from stationkeeping, drag compensation, attitude control and pointing, orbit raising and repositioning, to primary propulsion for interplanetary missions. Plasma thrusters have also been used as plasma sources in space for active space physics experiments and in the laboratory as plasma sources for reentry simulation, plasma processing and plasma injection in fusion devices. We will review research in the field, focusing on the plasma physics problems related to three classes of plasma thrusters: the Hall thruster (HT), the pulsed plasma thruster (PPT) and the magnetoplasmadynamic thruster (MPDT). The basic plasma acceleration and power loss mechanisms in each of these devices will be described along with the major plasma physics problems that control the thrust efficiency, stability and lifetime of these devices. We will review the recent advances and remaining questions relevant to the following important problems: macro and micro instabilities and turbulence, anomalous transport, ionization physics, plume divergence (HT and MPDT), current sheet dynamics and permeability (PPT).
NASA Astrophysics Data System (ADS)
Qi, Lei; Kwon, Jaemin; Hahm, T. S.; Jo, Gahyung
2016-06-01
Nonlinear bounce-averaged kinetic theory [B. H. Fong and T. S. Hahm, Phys. Plasmas 6, 188 (1999)] is used for magnetically trapped electron dynamics for the purpose of achieving efficient gyrokinetic simulations of Trapped Electron Mode (TEM) and Ion Temperature Gradient mode with trapped electrons (ITG-TEM) in shaped tokamak plasmas. The bounce-averaged kinetic equations are explicitly extended to shaped plasma equilibria from the previous ones for concentric circular plasmas, and implemented to a global nonlinear gyrokinetic code, Gyro-Kinetic Plasma Simulation Program (gKPSP) [J. M. Kwon et al., Nucl. Fusion 52, 013004 (2012)]. Verification of gKPSP with the bounce-averaged kinetic trapped electrons in shaped plasmas is successfully carried out for linear properties of the ITG-TEM mode and Rosenbluth-Hinton residual zonal flow [M. N. Rosenbluth and F. L. Hinton, Phys. Rev. Lett. 80, 724 (1998)]. Physics responsible for stabilizing effects of elongation on both ITG mode and TEM is identified using global gKPSP simulations. These can be understood in terms of magnetic flux expansion, leading to the effective temperature gradient R / L T ( 1 - E ') [P. Angelino et al., Phys. Rev. Lett. 102, 195002 (2009)] and poloidal wave length contraction at low field side, resulting in the effective poloidal wave number kθρi/κ.
Neoclassical Toroidal Viscosity Induced by Resonant Magnetic Perturbation in Tokamak Edge Plasma
NASA Astrophysics Data System (ADS)
Yan, Xing-Ting; Zhu, Ping; Sun, You-Wen
2015-11-01
In recent experiments, non-axisymmetric magnetic field perturbations due to external perturbations or plasma instabilities have been observed to strongly affect plasma rotation through neoclassical toroidal viscosity (NTV). In this work, we have calculated the NTV torque induced by resonant magnetic perturbation (RMP) in the edge plasma of a circular-shaped limiter tokamak, using the coupling of NIMROD and NTVTOK codes newly developed for this study. The resulting NTV torque is found to be sensitive to plasma β. In particular, when β is increased by two orders of magnitude, NTV torque is almost increased by ten orders of magnitude. The amplitude of NTV torque also depends on the toroidal mode number n of the plasma response to RMP. For a same amplitude of plasma response, the ion contribution to the resulting NTV torque increases with n, whereas the electron contribution decreases with n. This suggests the significance of nonlinear toroidal coupling in the generation of NTV torque, even when the RMP has only a single toroidal mode or helicity. Supported by National Magnetic Confinement Fusion Science Program of China Grant 2014GB124002.
NASA Astrophysics Data System (ADS)
Kim, Kimin; Ahn, J.-W.; Scotti, F.; Park, J.-K.; Menard, J. E.
2015-10-01
Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifies the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.
Plasma current start-up using the lower hybrid wave on the TST-2 spherical tokamak
NASA Astrophysics Data System (ADS)
Takase, Y.; Ejiri, A.; Inada, T.; Moeller, C. P.; Shinya, T.; Tsujii, N.; Yajima, S.; Furui, H.; Homma, H.; Imamura, K.; Nakamura, K.; Nakamura, K.; Sonehara, M.; Takeuchi, T.; Togashi, H.; Tsuda, S.; Yoshida, Y.
2015-12-01
Non-inductive plasma current start-up, ramp-up and sustainment by waves in the lower hybrid wave (LHW) frequency range at 200 MHz were investigated on the TST-2 spherical tokamak (R0 ≤ 0.38 m, a ≤ 0.25 m, Bt0 ≤ 0.3T, Ip ≤ 0.14 MA). Experimental results obtained using three types of antenna were compared. Both the highest plasma current (Ip = 18 kA) and the highest current drive figure of merit ηCD≡n¯eIpR0/PRF=1.4 ×1017 A/W/m2 were achieved using the capacitively-coupled combline (CCC) antenna, designed to excite the LHW with a sharp and highly directional wavenumber spectrum. For Ip greater than about 5 kA, high energy electrons accelerated by the LHW become the dominant carrier of plasma current. The low value of ηCD observed so far are believed to be caused by a rapid loss of energetic electrons and parasitic losses of the LHW energy in the plasma periphery. ηCD is expected to improve by an order of magnitude by increasing the plasma current to improve energetic electron confinement. In addition, edge power losses are expected to be reduced by increasing the toroidal magnetic field to improve wave accessibility to the plasma core, and by launching the LHW from the inboard upper region of the torus to achieve better single-pass absorption.
Kim, Kimin; Ahn, J. -W.; Scotti, F.; Park, J. -K.; Menard, J. E.
2015-09-03
Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifiesmore » the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Furthermore, amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.« less
Linking PFC surface characteristics and plasma performance in the Lithium Tokamak Experiment
NASA Astrophysics Data System (ADS)
Lucia, M.; Kaita, R.; Majeski, R.; Boyle, D. P.; Jaworski, M. A.; Schmitt, J. C.; Bedoya, F.; Allain, J. P.
2015-11-01
The Lithium Tokamak Experiment (LTX) is a spherical torus magnetic confinement device designed to accommodate lithium as the primary plasma-facing component (PFC). Results are presented from the implementation on LTX of the Materials Analysis and Particle Probe (MAPP), a compact in vacuo surface science diagnostic. With MAPP, in situ surface analysis techniques of x-ray photoelectron spectroscopy and thermal desorption spectroscopy are used to study evolution of the PFC surface chemistry in LTX as a function of varied lithium coating, hydrogen plasma exposure, and PFC surface temperature (20 - 300°C). Surface analysis results are then correlated with various measures of LTX plasma performance, including toroidal plasma current, line-integrated plasma density, and density-normalized impurity emission. Lithium coatings are observed to convert within hours to Li2O by gettering oxygen from both the residual vacuum and the PFC substrate. However, plasma performance remains elevated even with discharges operating against Li2O -coated PFCs. Hydrogen is retained by these Li2O coatings during a discharge, but it is almost completely desorbed as outgassed H2 in the minutes following the discharge; no persistent LiH formation is observed. This work was supported by U.S. DOE contracts DE-AC02-09CH11466, DE-AC52-07NA27344, and DE-SC0010717, as well as by an NSF GRFP fellowship under grant DGE-0646086.
A penalization technique to model plasma facing components in a tokamak with temperature variations
Paredes, A.; Bufferand, H.; Ciraolo, G.; Schwander, F.; Serre, E.; Ghendrih, P.; Tamain, P.
2014-10-01
To properly address turbulent transport in the edge plasma region of a tokamak, it is mandatory to describe the particle and heat outflow on wall components, using an accurate representation of the wall geometry. This is challenging for many plasma transport codes, which use a structured mesh with one coordinate aligned with magnetic surfaces. We propose here a penalization technique that allows modeling of particle and heat transport using such structured mesh, while also accounting for geometrically complex plasma-facing components. Solid obstacles are considered as particle and momentum sinks whereas ionic and electronic temperature gradients are imposed on both sides of the obstacles along the magnetic field direction using delta functions (Dirac). Solutions exhibit plasma velocities (M=1) and temperatures fluxes at the plasma–wall boundaries that match with boundary conditions usually implemented in fluid codes. Grid convergence and error estimates are found to be in agreement with theoretical results obtained for neutral fluid conservation equations. The capability of the penalization technique is illustrated by introducing the non-collisional plasma region expected by the kinetic theory in the immediate vicinity of the interface, that is impossible when considering fluid boundary conditions. Axisymmetric numerical simulations show the efficiency of the method to investigate the large-scale transport at the plasma edge including the separatrix and in realistic complex geometries while keeping a simple structured grid.
Clarification of symmetry breaking mechanism in intrinsic rotation of tokamak plasmas
NASA Astrophysics Data System (ADS)
Yi, S.; Kwon, J. M.; Rhee, T.; Diamond, P. H.; Kim, J. Y.
2010-11-01
Intrinsic rotation of tokamak plasmas is considered to be generated by non-diffusive stress (i.e. residual stress) induced by asymmetric k|| turbulence spectrum. To study the symmetry breaking mechanisms in intrinsic rotation, we have performed numerical simulations of intrinsic rotation by ITG turbulence using the gKPSP code, a delta-f global PIC code for tokamak. It is found that not only distortion of turbulence spectrum by ExB shear but also spatial diffusion of wave momentum driven by turbulence intensity gradient play an important role in the symmetry breaking mechanism, as expected from a theory [1]. It is hard to recognize individual contribution of ExB shear and turbulence intensity gradient to the residual stress because their evolution is strongly coupled with the prey-predator feature [2]. To clarify their role, a comprehensive analysis including their nonlinear coupling is performed. The key symmetry breaking mechanism is identified for various physics situations. [4pt] [1] P.H. Diamond, et al., Phys. of Plasmas 15, 012303 (2008). [0pt] [2] P.H. Diamond, et al., PRL 72, 2565 (1994).
The National Spherical Tokamak Experiment at the Princeton Plasma Physics Laboratory
1995-12-01
The Department of Energy (DOE) has prepared an Environmental Assessment (EA), DOE/EA-1108, evaluating the environmental effects of the proposed construction and operation of the National Spherical Tokamak Experiment (NSTX) within the existing C-Stellarator (CS) Building at the Princeton Plasma Physics Laboratory, Princeton, New Jersey. The purpose of the NSTX is to investigate the physics of spherically shaped plasmas as an alternative path to conventional tokamaks for development of fusion energy. Fusion energy has the potential to help compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Construction of the NSTX in the CS Building would require the dismantling and removal of the existing unused Princeton Large Torus (PLT) device, part of which would be reused to construct the NSTX. Based on the analyses in the EA, the DOE has determined that the proposed action does not constitute a major federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act (NEPA) of 1969, 42 U.S.C. 4,321 et seq. The preparation of an Environmental Impact Statement is not required. Thus, the DOE is issuing a FONSI pursuant to the Council on Environmental Quality regulations implementing NEPA (40 CFR Parts 1500--1508) and the DOE NEPA implementing regulations (10 CFR Part 1021).
Modeling of runaway electron damage for the design of tokamak plasma facing components
NASA Astrophysics Data System (ADS)
Niemer, K. A.; Gilligan, J. G.; Croessmann, C. D.; Bolt, H. H.
1990-04-01
Cracking, craters, spotty damage (discoloration), and missing chunks of material have been observed on limiters and along the midplane of tokamak inner walls. This damage is assumed to be due to runaway electron discharges. These runaway electrons have been predicted to range in energy from a few MeV to several hundred MeV. The energy density from the runaway electron discharges ranges from 10 to 500 MJ/sq m over pulse lengths of 5 to 50 msec. The PTA code package is a three dimensional, time dependent, computational code package used to predict energy deposition, temperature rise, and damage on tokamak first wall and limiter materials form runaway electron impact. Two experiments were modeled to validate the PTA code package. The first experiment tested the thermal and structural response from high energy electron impact on different fusion materials, and the second experiment simulated runaway electrons scattering through a plasma facing surface (graphite) into an internal structure (copper). The PTA calculations compared favorably with the experimental results. In particular, the PTA models identified gap conductance, thermal contact, x ray generation in materials, and the placement of high stopping power materials as key factors in the design of plasma facing components, resistant to runaway electron damage.
TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry
X. Q. Xu; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.
2010-05-28
We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.
TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry
X. Q. Xu; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.
2010-05-28
We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less
TEMPEST simulations of the plasma transport in a single-null tokamak geometry
NASA Astrophysics Data System (ADS)
Xu, X. Q.; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.
2010-06-01
We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.
Recent Advances in Plasma Acceleration
Hogan, Mark
2007-03-19
The costs and the time scales of colliders intended to reach the energy frontier are such that it is important to explore new methods of accelerating particles to high energies. Plasma-based accelerators are particularly attractive because they are capable of producing accelerating fields that are orders of magnitude larger than those used in conventional colliders. In these accelerators a drive beam, either laser or particle, produces a plasma wave (wakefield) that accelerates charged particles. The ultimate utility of plasma accelerators will depend on sustaining ultra-high accelerating fields over a substantial length to achieve a significant energy gain. More than 42 GeV energy gain was achieved in an 85 cm long plasma wakefield accelerator driven by a 42 GeV electron drive beam in the Final Focus Test Beam (FFTB) Facility at SLAC. Most of the beam electrons lose energy to the plasma wave, but some electrons in the back of the same beam pulse are accelerated with a field of {approx}52 GV/m. This effectively doubles their energy, producing the energy gain of the 3 km long SLAC accelerator in less than a meter for a small fraction of the electrons in the injected bunch. Prospects for a drive-witness bunch configuration and high-gradient positron acceleration experiments planned for the SABER facility will be discussed.
Particle pinch and collisionality in gyrokinetic simulations of tokamak plasma turbulence
Angioni, C.; Candy, J.; Waltz, R. E.; Fable, E.; Maslov, M.; Weisen, H.; Peeters, A. G.
2009-06-15
The generic problem of how, in a turbulent plasma, the experimentally relevant conditions of a particle flux very close to the null are achieved, despite the presence of strong heat fluxes, is addressed. Nonlinear gyrokinetic simulations of plasma turbulence in tokamaks reveal a complex dependence of the particle flux as a function of the turbulent spatial scale and of the velocity space as collisionality is increased. At experimental values of collisionality, the particle flux is found close to the null, in agreement with the experiment, due to the balance between inward and outward contributions at small and large scales, respectively. These simulations provide full theoretical support to the prediction of a peaked density profile in a future nuclear fusion reactor.
Measurement of sheared flows in the edge plasma of the CASTOR tokamak
Brotankova, J.; Stockel, J.; Seidl, J.; Duran, I.; Hron, M.
2009-11-15
The ion saturation current and floating potential are measured with high temporal (1 {mu}s) and spatial (2.5 mm) resolutions at the plasma edge of the CASTOR tokamak by two poloidally spaced radial arrays of Langmuir probes. The radial electric field and the phase velocity of plasma fluctuations are estimated. The position of the velocity shear layer (VSL) is localized with a high precision. The shearing rate {omega}{sub ExB} determined and found to be comparable with the inverse of the correlation time of fluctuations 1/{tau}{sub ac} outside the VSL and about five times higher in the proximity of the VSL. A small impact of the shear on fluctuation level at the VSL is observed also in the statistic parameters of the U{sub fl} and I{sub sat}.
Measurement of the edge plasma rotation on J-TEXT tokamak
Cheng, Z. F.; Luo, J.; Wang, Z. J.; Zhang, Z. P.; Zhang, X. L.; Hou, S. Y.; Cheng, C.; Zhuang, G.
2013-07-15
A multi-channel high resolution spectrometer was developed for the measurement of the edge plasma rotation on J-TEXT tokamak. With the design of two opposite viewing directions, the poloidal and toroidal rotations can be measured simultaneously, and velocity accuracy is up to 1 km/s. The photon flux was enhanced by utilizing combined optical fiber. With this design, the time resolution reaches 3 ms. An assistant software “Spectra Assist” was developed for implementing the spectrometer control and data analysis automatically. A multi-channel monochromatic analyzer is designed to get the location of chosen ions simultaneously through the inversion analysis. Some preliminary experimental results about influence of plasma density, different magnetohydrodynamics behaviors, and applying of biased electrode are presented.
Generation of a magnetic island by edge turbulence in tokamak plasmas
Poyé, A.; Agullo, O.; Muraglia, M.; Benkadda, S.; Dubuit, N.; Garbet, X.; Sen, A.
2015-03-15
We investigate, through extensive 3D magneto-hydro-dynamics numerical simulations, the nonlinear excitation of a large scale magnetic island and its dynamical properties due to the presence of small-scale turbulence. Turbulence is induced by a steep pressure gradient in the edge region [B. D. Scott, Plasma Phys. Controlled Fusion 49, S25 (2007)], close to the separatrix in tokamaks where there is an X-point magnetic configuration. We find that quasi-resonant localized interchange modes at the plasma edge can beat together and produce extended modes that transfer energy to the lowest order resonant surface in an inner stable zone and induce a seed magnetic island. The island width displays high frequency fluctuations that are associated with the fluctuating nature of the energy transfer process from the turbulence, while its mean size is controlled by the magnetic energy content of the turbulence.
Plasma radiometry with 30 chord resolution for fast transients in the DIII-D tokamak
Gray, D.S.; Hollmann, E.M.; Luckhardt, S.C.; Chalfant, J.; Chousal, L.; Hernandez, R.; Jones, E.; Kellman, A.G.
2004-10-01
A diagnostic capable of providing time resolved measurements of plasma radiated power during disruptions and other fast transients, e.g., edge localized modes has been employed in the DIII-D tokamak. The radiation is detected with absolute extreme ultraviolet (AXUV) photodiode arrays. Thirty chords from a single port provide measurements from a full slice of the plasma at one toroidal location. The analog bandwidth is up to 1 MHz for the brightest events, i.e., disruptions. Active cooling of the diode arrays prevents damage during high temperature vessel baking. Effective responsivity values of 0.12-0.18 A/W are taken from previous work on the application of AXUV diodes in DIII-D. The total radiated energy in disruptions typically agrees with bolometer measurements within about 12%.
Kinetic toroidal Alfv{acute e}n eigenmodes in finite-{beta} tokamak plasmas
Zheng, L.-.; Chen, L.
1998-04-01
Kinetic toroidal Alfv{acute e}n eigenmodes (KTAEs) in finite-{beta} circular tokamak plasmas are investigated. Here, {beta} is the ratio between plasma and magnetic pressures, and, formally, {beta}{approximately}scr(O)(r/Rq{sup 2}), with q being the safety factor, r and R denoting, respectively, minor and major radii. A new effect associated with finite parallel electric field effect due to the ion magnetic drift (IMD) is discovered, which is of the same order as that due to the well-known effect associated with the finite ion Larmor radii. The IMD-induced parallel electric field effect is shown to contribute to potential wells for the eigenmodes. Therefore, the IMD-induced parallel electric field effect can discretize the lower Alfv{acute e}n continuum along the real eigenfrequency. Subsequently, this new branch of KTAEs could be readily destabilized by the energetic ions. {copyright} {ital 1998 American Institute of Physics.}
Yamada, M.; Manickam, J.; Pomphrey, N.; Levinton, F.M.; Nagayama, Y.
1994-01-01
Magnetic reconnection is investigated in high temperature TFTR tokamak plasmas by a set of non-perturbative diagnostics. During the crash phase of sawtooth oscillations in the plasma discharges, the ECE (electron cyclotron emission) diagnostic measures a fast flattening of the 2-D electron temperature profile in a poloidal plane, an observation consistent with the Kadomtsev reconnection theory. On the other hand motional Stark effect(MSE) measurements indicate that central q values do not relax to unity after the crash, but increase only by 5-10%, typically from 0.7 to 0.75. The latter result is in contradiction with the models of Kadomtsev and/or Wesson. A heuristic model for the magnetic reconnection at the sawtooth crash is also presented.
Measurement of the edge plasma rotation on J-TEXT tokamak.
Cheng, Z F; Luo, J; Wang, Z J; Zhang, Z P; Zhang, X L; Hou, S Y; Cheng, C; Zhuang, G
2013-07-01
A multi-channel high resolution spectrometer was developed for the measurement of the edge plasma rotation on J-TEXT tokamak. With the design of two opposite viewing directions, the poloidal and toroidal rotations can be measured simultaneously, and velocity accuracy is up to 1 km∕s. The photon flux was enhanced by utilizing combined optical fiber. With this design, the time resolution reaches 3 ms. An assistant software "Spectra Assist" was developed for implementing the spectrometer control and data analysis automatically. A multi-channel monochromatic analyzer is designed to get the location of chosen ions simultaneously through the inversion analysis. Some preliminary experimental results about influence of plasma density, different magnetohydrodynamics behaviors, and applying of biased electrode are presented. PMID:23902064
Shielding of External Magnetic Perturbations By Torque In Rotating Tokamak Plasmas
Park, Jong-Kyu; Boozer, Allen H.; Menard, Jonathan E.; Gerhardt, Stefan P.; Sabbagh, Steve A.
2009-08-24
The imposition of a nonaxisymmetric magnetic perturbation on a rotating tokamak plasma requires energy and toroidal torque. Fundamental electrodynamics implies that the torque is essentially limited and must be consistent with the external response of a plasma equilibrium ƒ = j x B. Here magnetic measurements on National Spherical Torus eXperiment (NSTX) device are used to derive the energy and the torque, and these empirical evaluations are compared with theoretical calculations based on perturbed scalar pressure equilibria ƒ = ∇p coupled with the theory of nonambipolar transport. The measurement and the theory are consistent within acceptable uncertainties, but can be largely inconsistent when the torque is comparable to the energy. This is expected since the currents associated with the torque are ignored in scalar pressure equilibria, but these currents tend to shield the perturbation.
Modification of plasma rotation with resonant magnetic perturbations in the STOR-M tokamak
NASA Astrophysics Data System (ADS)
Elgriw, S.; Liu, Y.; Hirose, A.; Xiao, C.
2016-04-01
The toroidal plasma flow velocity of impurity ions has been significantly modified in the Saskatchewan Torus-Modified (STOR-M) tokamak by means of resonant magnetic perturbations (RMP). It has been found that the toroidal flow velocities of OV and CVI impurity ions change towards the co-current direction after the application of a current through a set of (l = 2, n = 1) RMP field coils. It has been observed that the reduction of the toroidal flow velocity is closely correlated to the reduction of the magnetohydrodynamic (MHD) fluctuation frequency measured by Mirnov coils. Modulation of the flow velocity has been achieved by switching the RMP current pulses. Non-resonant magnetic perturbations have also induced a much smaller change in the toroidal plasma flow. A theoretical model has been adopted to assess the contributions of different drift mechanisms to magnetic islands rotation in STOR-M.
Multidirectional plasma flow measurement by Gundestrup Probe in scrape-off layer of ADITYA tokamak
Sangwan, Deepak; Jha, Ratneshwar; Tanna, Rakesh L.
2015-11-15
Multidirectional plasma flow measurements by using Gundestrup Probe in the scrape-off layer of ADITYA tokamak are presented. The ADITYA Gundestrup Probe-head consists of eight plates arranged around the ceramic rod and three pins normal to side plates. Plates are used to measure both parallel and perpendicular flows simultaneously and pins are used to measure plasma density and floating potential. A comparison of direct perpendicular flow measurement and by two other plates of Gundestrup Probe is presented. Possible causes of perpendicular flows are identified and compared with the measured flows. It is observed that the mechanism of the parallel flow and the perpendicular flow is different only at high parallel Mach number. A puff of the working gas is used to study its effect on the perpendicular flows and its reversal with the gas puff is observed.
The Construction of Plasma Density Feedback Control System on J-TEXT Tokamak
NASA Astrophysics Data System (ADS)
Ke, Xin; Chen, Zhipeng; Ba, Weigang; Shu, Shuangbao; Gao, Li; Zhang, Ming; Zhuang, Ge
2016-02-01
The plasma density feedback control system (PDFCS) has been established on the Joint Texas Experimental Tokamak (J-TEXT) for meeting the need for an accurate plasma density in physical experiments. It consists of a density measurement subsystem, a feedback control subsystem and a gas puffing subsystem. According to the characteristic of the gas puffing system, a voltage amplitude control mode has been applied in the feedback control strategy, which is accomplished by the proportion, integral and differential (PID) controller. In this system, the quantity calibration of gas injection, adjusted responding to the change of the density signal, has been carried out. Some experimental results are shown and discussed. supported by the National Magnetic Confinement Fusion Science Program (Nos. 2014GB103001 and 2013GB106001) and National Natural Science Foundation of China (Nos. 11305070 and 11105028)
Functional form for plasma velocity in a rapidly rotating tokamak discharge
Burrell, K. H.; Chrystal, C.
2014-07-15
A recently developed technique using charge exchange spectroscopy determines the ion poloidal rotation in tokamak plasmas from the poloidal variation in the toroidal angular rotation speed. The basis for this technique is the functional form for the plasma velocity calculated from the equilibrium equations. The initial development of this technique utilized the functional form determined for conditions where the ion toroidal rotation speed is much smaller than the ion thermal speed. There are cases, however, where the toroidal rotation can be comparable to the ion thermal speed, especially for high atomic number impurities. The present paper extends the previous analysis to this high rotation speed case and demonstrates how to extract the poloidal rotation speed from measurements of the toroidal angular rotation speed at two points on a flux surface.
Cleaning of HT-7 Tokamak Exposed First Mirrors by Radio Frequency Magnetron Sputtering Plasma
NASA Astrophysics Data System (ADS)
Yan, Rong; Chen, Junling; Chen, Longwei; Ding, Rui; Zhu, Dahuan
2014-12-01
The stainless steel (SS) first mirror pre-exposed in the deposition-dominated environment of the HT-7 tokamak was cleaned in the newly built radio frequency (RF) magnetron sputtering plasma device. The deposition layer on the FM surface formed during the exposure was successfully removed by argon plasma with a RF power of about 80 W and a gas pressure of 0.087 Pa for 30 min. The total reflectivity of the mirrors was recovered up to 90% in the wavelength range of 300-800 nm, while the diffuse reflectivity showed a little increase, which was attributed to the increase of surface roughness in sputtering, and residual contaminants. The FMs made from single crystal materials could help to achieve a desired recovery of specular reflectivity in the future.
Henriques, R. B. Malaquias, A.; Nedzelskiy, I. S.; Silva, C.; Coelho, R.; Figueiredo, H.; Fernandes, H.
2014-11-15
The Heavy Ion Beam Diagnostic (HIBD) on the tokamak ISTTOK (Instituto Superior Técnico TOKamak) has been modified, in terms of signal conditioning, to measure the local fluctuations of the n{sub e}σ{sub 1,2}(T{sub e}) product (plasma density times the effective ionization cross-section) along the tokamak minor diameter, in 12 sample volumes in the range of −0.7a < r < 0.7a, with a maximum delay time of 1 μs. The corresponding signals show high correlation with the magnetic Mirnov coils in the characteristic MHD frequency range of ISTTOK plasmas and enable the identification of tearing modes. This paper describes the HIBD signal conditioning system and presents a preliminary analysis of the radial profile measurements of local n{sub e}σ{sub 1,2}(T{sub e}) fluctuations.
Ou, Jing; Wu, Guojiang; Li, Xinxia
2014-07-15
Distribution of the intrinsic rotation due to collisionless ion orbit loss near the tokamak edge region is studied by using an analytical model based on ion guiding center orbit approximation. A peak of the averaged ion orbit loss momentum fraction is found very near inside the separatrix region in a double null divertor configuration but is not found inside the last closed flux surface region in an outer limiter configuration. For the double null divertor configuration, the intrinsic rotation due to ion orbit loss depends on the plasma shape. With the increase in elongation and triangularity, the peak of the averaged ion orbit loss momentum fraction increases and it moves inward for the lower plasma current.
NASA Astrophysics Data System (ADS)
Kim, YoungJin; Yoo, Min-Gu; Kim, S. H.; Na, Yong-Su
2015-01-01
A field-based new adaptive mesh generator, VEGA (VEctor-following Grid generator for Adaptive mesh), is developed for 2-D core-edge coupled tokamak plasma transport simulations. VEGA can generate time-varying and spatially non-uniform grids by using a stretching function. It provides two operation modes for generating non-uniform radial distributions. One is so-called ion mode where the grid is automatically generated by considering the ion temperature gradient which plays an important role in the ion and the momentum transport mechanism of a tokamak plasma. The other is so-called high-gradient mode where the grid is produced by considering the locality of plasma profiles which appears particularly in transport barriers. VEGA is benchmarked with a conventional code for a reference double null (DN) KSTAR divertor configuration. Three factors are newly introduced in this work to evaluate the quality of a grid. It is found that VEGA is particularly suitable for delicate integrated simulations of the plasma edge and the scrape off layer (SOL) due to its high cell orthogonality and low radial flux deviation. Quality of non-uniform grids generated by the two operation modes of VEGA, the ion mode and the high-gradient mode is examined. A more refined grid is found near the edge region characterized with steeper gradients whereas coarser one in the core region. Such fine grids at the edge region can result in highly reduced radial flux deviation, which is indeed important for analysis of edge-SOL physics with time-varying simulations.
Zhu, Y. B. Liu, D.; Heidbrink, W. W.; Zhang, J. Z.; Qi, M. Z.; Xia, S. B.; Wan, B. N.; Li, J. G.
2014-11-15
Full function integrated, compact silicon photodiode based solid state neutral particle analyzers (ssNPA) have been developed for energetic particle (EP) relevant studies on the Experimental Advanced Superconducting Tokamak (EAST). The ssNPAs will be mostly operated in advanced current mode with a few channels to be operated in conventional pulse-counting mode, aiming to simultaneously achieve individually proved ultra-fast temporal, spatial, and spectral resolution capabilities. The design details together with considerations on EAST specific engineering realities and physics requirements are presented. The system, including a group of single detectors on two vertical ports and two 16-channel arrays on a horizontal port, can provide both active and passive charge exchange measurements. ssNPA detectors, with variable thickness of ultra thin tungsten dominated foils directly deposited on the front surface, are specially fabricated and utilized to achieve about 22 keV energy resolution for deuterium particle detection.
Advances in cold plasma technology
Technology Transfer Automated Retrieval System (TEKTRAN)
Foodborne pathogens continue to be an issue on a variety of commodities, prompting research into novel interventions. Cold plasma is a nonthermal food processing technology which uses energetic, reactive gases to inactivate contaminating microbes on meats, poultry and fruits and vegetables. The prim...
Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas
NASA Astrophysics Data System (ADS)
Horacek, J.; Pitts, R. A.; Adamek, J.; Arnoux, G.; Bak, J.-G.; Brezinsek, S.; Dimitrova, M.; Goldston, R. J.; Gunn, J. P.; Havlicek, J.; Hong, S.-H.; Janky, F.; LaBombard, B.; Marsen, S.; Maddaluno, G.; Nie, L.; Pericoli, V.; Popov, Tsv; Panek, R.; Rudakov, D.; Seidl, J.; Seo, D. S.; Shimada, M.; Silva, C.; Stangeby, P. C.; Viola, B.; Vondracek, P.; Wang, H.; Xu, G. S.; Xu, Y.; Contributors, JET
2016-07-01
As in many of today’s tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, {{q}||} in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as {{q}||}={{q}0}\\text{exp} ~≤ft(-r/λ q\\text{omp}\\right) , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, λ q\\text{omp} . The initial choice of λ q\\text{omp} , which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with R=\\text{0}\\text{.4--2}\\text{.8} \\text{m}, {{B}0}=\\text{1}\\text{.2--7}\\text{.5} \\text{T}, {{I}\\text{p}}=\\text{9--2500} \\text{kA}. Measurements of λ q\\text{omp} in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar
Investigation of magnetic reconnection during a sawtooth crash in a high-temperature tokamak plasma
NASA Astrophysics Data System (ADS)
Yamada, M.; Levinton, F. M.; Pomphrey, N.; Budny, R.; Manickam, J.; Nagayama, Y.
1994-10-01
In this paper a laboratory investigation is made on magnetic reconnection in high-temperature Tokamak Fusion Test Reactor (TFTR) plasmas [Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 51]. The motional Stark effect (MSE) diagnostic is employed to measure the pitch angle profile of magnetic field lines, and hence the q profile. An analytical expression that relates pitch angle to q profile is presented for a toroidal plasma with circular cross section. During the crash phase of sawtooth oscillations in plasma discharges, the ECE (electron cyclotron emission) diagnostic measures a fast flattening of the two-dimensional (2-D) electron temperature profile in a poloidal plane, an observation consistent with the Kadomtsev reconnection theory. On the other hand, the MSE measurements indicate that central q values do not relax to unity after the crash, but increase only by 5%-15%, typically from 0.7 to 0.8. The latter result is in contradiction with the 2-D models of Kadomtsev and/or Wesson. In the present study this puzzle is addressed by a simultaneous analysis of electron temperature and q profile evolutions. Based on a heuristic model for magnetic reconnection during the sawtooth crash, the small change of q, i.e., partial reconnection, is attributed to the precipitous drop of pressure gradients that drive the instability and the reconnection process, as well as flux conserving plasma dynamics.
Investigation of magnetic reconnection during a sawtooth crash in a high-temperature tokamak plasma
Yamada, M.; Levinton, F.M.; Pomphrey, N.; Budny, R.; Manickam, J.; Nagayama, Y. )
1994-10-01
In this paper a laboratory investigation is made on magnetic reconnection in high-temperature Tokamak Fusion Test Reactor (TFTR) plasmas [[ital Plasma] [ital Physics] [ital and] [ital Controlled] [ital Nuclear] [ital Fusion] [ital Research] 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 51]. The motional Stark effect (MSE) diagnostic is employed to measure the pitch angle profile of magnetic field lines, and hence the [ital q] profile. An analytical expression that relates pitch angle to [ital q] profile is presented for a toroidal plasma with circular cross section. During the crash phase of sawtooth oscillations in plasma discharges, the ECE (electron cyclotron emission) diagnostic measures a fast flattening of the two-dimensional (2-D) electron temperature profile in a poloidal plane, an observation consistent with the Kadomtsev reconnection theory. On the other hand, the MSE measurements indicate that central [ital q] values do not relax to unity after the crash, but increase only by 5%--15%, typically from 0.7 to 0.8. The latter result is in contradiction with the 2-D models of Kadomtsev and/or Wesson. In the present study this puzzle is addressed by a simultaneous analysis of electron temperature and [ital q] profile evolutions. Based on a heuristic model for magnetic reconnection during the sawtooth crash, the small change of [ital q], i.e., partial reconnection, is attributed to the precipitous drop of pressure gradients that drive the instability and the reconnection process, as well as flux conserving plasma dynamics.
Analytical modeling of equilibrium of strongly anisotropic plasma in tokamaks and stellarators
Lepikhin, N. D.; Pustovitov, V. D.
2013-08-15
Theoretical analysis of equilibrium of anisotropic plasma in tokamaks and stellarators is presented. The anisotropy is assumed strong, which includes the cases with essentially nonuniform distributions of plasma pressure on magnetic surfaces. Such distributions can arise at neutral beam injection or at ion cyclotron resonance heating. Then the known generalizations of the standard theory of plasma equilibrium that treat p{sub ‖} and p{sub ⊥} (parallel and perpendicular plasma pressures) as almost constant on magnetic surfaces are not applicable anymore. Explicit analytical prescriptions of the profiles of p{sub ‖} and p{sub ⊥} are proposed that allow modeling of the anisotropic plasma equilibrium even with large ratios of p{sub ‖}/p{sub ⊥} or p{sub ⊥}/p{sub ‖}. A method for deriving the equation for the Shafranov shift is proposed that does not require introduction of the flux coordinates and calculation of the metric tensor. It is shown that for p{sub ⊥} with nonuniformity described by a single poloidal harmonic, the equation for the Shafranov shift coincides with a known one derived earlier for almost constant p{sub ⊥} on a magnetic surface. This does not happen in the other more complex case.
Dynamics of tokamak plasma surface current in 3D ideal MHD model
NASA Astrophysics Data System (ADS)
Galkin, Sergei A.; Svidzinski, V. A.; Zakharov, L. E.
2013-10-01
Interest in the surface current which can arise on perturbed sharp plasma vacuum interface in tokamaks was recently generated by a few papers (see and references therein). In dangerous disruption events with plasma-touching-wall scenarios, the surface current can be shared with the wall leading to the strong, damaging forces acting on the wall A relatively simple analytic definition of δ-function surface current proportional to a jump of tangential component of magnetic field nevertheless leads to a complex computational problem on the moving plasma-vacuum interface, requiring the incorporation of non-linear 3D plasma dynamics even in one-fluid ideal MHD. The Disruption Simulation Code (DSC), which had recently been developed in a fully 3D toroidal geometry with adaptation to the moving plasma boundary, is an appropriate tool for accurate self-consistent δfunction surface current calculation. Progress on the DSC-3D development will be presented. Self-consistent surface current calculation under non-linear dynamics of low m kink mode and VDE will be discussed. Work is supported by the US DOE SBIR grant #DE-SC0004487.
Impact of plasma core profiles on MHD stability at tokamak edge pedestal
NASA Astrophysics Data System (ADS)
Aiba, N.; Urano, H.
2014-11-01
Impact of plasma core profiles on magnetohydrodynamics (MHD) stability at tokamak edge pedestal is investigated numerically to extend an operation regime for small amplitude grassy edge localized mode (ELM). With the hypotheses that pedestal pressure profile can be predicted with the EPED1 model and the trigger of grassy ELM is an ideal ballooning mode, the impacts of plasma poloidal beta and plasma internal inductance on edge MHD stability are investigated, the parameters of which are related to plasma core profiles and are important parameters for grassy ELMy H-modes in JET quasi-double null plasma. The numerical results indicate that a ballooning mode can be destabilized by decreasing poloidal beta and/or internal inductance. In contrast, it is confirmed that pedestal density, which is also an important parameter for realizing grassy ELMy H-mode, can stabilize a ballooning mode. In combination with these trends, it is possible to relax the necessary conditions for grassy ELMy H-mode by adjusting the parameters carefully, though this relaxation destabilizes type-I ELM more easily due to the increase in edge current density.
NASA Astrophysics Data System (ADS)
Lundberg, D. P.; Granstedt, E.; Kaita, R.; Majeski, R.
2011-10-01
Lithium Tokamak Experiment (LTX) plasmas with lithium-coated walls have demonstrated low-recycling conditions, with substantially higher fueling requirements and reductions in edge neutral emission. Most fueling systems, such as wall-mounted gas puffers or supersonic gas injectors, are ill-suited for use with low-recycling plasmas, as they primarily source low-density gas into the plasma edge. A Molecular Cluster Injector (MCI) has been installed to improve fueling efficiency by increasing the penetration of neutrals into the plasma core. The MCI molecular density has been measured with an electron beam, with nH2exceeding 1016cm-3 more than 15cm from the nozzle. These densities are 100-1000 the LTX ne, making the MCI suitable for testing high-density fueling. By varying the MCI pressure, temperature, and location relative to the plasma, the relative importance of the molecular density and the degree of cluster formation within the supersonic jet can be studied. The effects of MCI fueling on LTX ne profiles is discussed. Supported by DOE contract number DE-AC02-09CH11466.
Simulations of the L-H transition on experimental advanced superconducting Tokamak
Weiland, Jan
2014-12-15
We have simulated the L-H transition on the EAST tokamak [Baonian Wan, EAST and HT-7 Teams, and International Collaborators, “Recent experiments in the EAST and HT-7 superconducting tokamaks,” Nucl. Fusion 49, 104011 (2009)] using a predictive transport code where ion and electron temperatures, electron density, and poloidal and toroidal momenta are simulated self consistently. This is, as far as we know, the first theory based simulation of an L-H transition including the whole radius and not making any assumptions about where the barrier should be formed. Another remarkable feature is that we get H-mode gradients in agreement with the α – α{sub d} diagram of Rogers et al. [Phys. Rev. Lett. 81, 4396 (1998)]. Then, the feedback loop emerging from the simulations means that the L-H power threshold increases with the temperature at the separatrix. This is a main feature of the C-mod experiments [Hubbard et al., Phys. Plasmas 14, 056109 (2007)]. This is also why the power threshold depends on the direction of the grad B drift in the scrape off layer and also why the power threshold increases with the magnetic field. A further significant general H-mode feature is that the density is much flatter in H-mode than in L-mode.
Wahlberg, C.
2009-11-15
Analytical theory and two different magnetohydrodynamical stability codes are used in a study of the effects of toroidal plasma rotation on the stability of the ideal, internal kink mode in tokamaks. The focus of the paper is on the role that the centrifugal effects on the plasma equilibrium play for the stability of this mode, and results from one code where centrifugal effects are self-consistently included (CASTOR-FLOW) [E. Strumberger et al., Nucl. Fusion 45, 1156 (2005)] are compared with the results from another code where such effects are not taken into account (MISHKA-F) [I. T. Chapman et al., Phys. Plasmas 13, 062511 (2006)]. It is found that, even at rather modest flow speeds, the centrifugal effects are very important for the stability of the internal kink mode. While the results from the two codes can be quite similar for certain profiles in the plasma, completely opposite results are obtained for other profiles. A very good agreement between analytical theory and the numerical results are, both for inconsistent and consistent equilibria, found for plasmas with large aspect ratio. From the analytical theory, the distinctly different stability properties of equilibria with and without centrifugal effects included can be traced to the stabilizing effect of the geodesic acoustic mode (GAM) induced by the plasma rotation. This GAM exists solely as a consequence of the nonuniform plasma density and pressure created by the centrifugal force on the flux surfaces, and a stabilizing coupling of the internal kink instability to this mode cannot therefore take place if the centrifugal effects are not included in the equilibrium. In addition to the GAM stabilization, the effects of the radial profiles of the plasma density and rotation velocity are also found to be significant, and the importance of these effects increases with decreasing aspect ratio.
Novel approaches for mitigating runaway electrons and plasma disruptions in ADITYA tokamak
NASA Astrophysics Data System (ADS)
Tanna, R. L.; Ghosh, J.; Chattopadhyay, P. K.; Dhyani, Pravesh; Purohit, Shishir; Joisa, S.; Rao, C. V. S.; Panchal, V. K.; Raju, D.; Jadeja, K. A.; Bhatt, S. B.; Gupta, C. N.; Chavda, Chhaya; Kulkarni, S. V.; Shukla, B. K.; Praveenlal E., V.; Raval, Jayesh; Amardas, A.; Atrey, P. K.; Dhobi, U.; Manchanda, R.; Ramaiya, N.; Patel, N.; Chowdhuri, M. B.; Jha, S. K.; Jha, R.; Sen, A.; Saxena, Y. C.; Bora, D.; the ADITYA Team
2015-06-01
This paper summarizes the results of recent dedicated experiments on disruption control and runaway mitigation carried out in ADITYA, which are of the utmost importance for the successful operation of large size tokamaks, such as ITER. It is quite a well-known fact that disruptions in tokamaks must be avoided. Disruptions, induced by hydrogen gas puffing, are successfully avoided by two innovative techniques in ADITYA using a bias electrode placed inside the last closed flux surface and applying an ion cyclotron resonance pulse with a power of ∼50 to 70 kW. These experiments led to better understanding of the disruption avoidance mechanisms and also can be thought of as one of the options for disruption avoidance in ITER. In both cases, the physical mechanism seems to be the control of magnetohydrodynamic modes due to increased poloidal rotation of edge plasma generated by induced radial electric fields. Real time avoidance of disruption with identifying proper precursors in both the mechanisms is successfully attempted. Further, analysing thoroughly the huge database of different types of spontaneous and deliberately-triggered disruptions from ADITYA, a significant contribution has been made to the international disruption database (ITPA). Furthermore, the mitigation of the runaway electron generated mainly during disruptions remains a challenging topic in present tokamak research as these high-energy electrons can cause severe damage to in-vessel components and the vacuum vessel. A simple technique has been implemented in ADITYA to mitigate the runaway electrons before they can gain high energies using a localized vertical magnetic field perturbation applied at one toroidal location to extract runaway electrons.
Advances and Challenges in Computational Plasma Science
W.M. Tang; V.S. Chan
2005-01-03
Scientific simulation, which provides a natural bridge between theory and experiment, is an essential tool for understanding complex plasma behavior. Recent advances in simulations of magnetically-confined plasmas are reviewed in this paper with illustrative examples chosen from associated research areas such as microturbulence, magnetohydrodynamics, and other topics. Progress has been stimulated in particular by the exponential growth of computer speed along with significant improvements in computer technology.
Experiments on Turbulence and Transport in the Edge Plasma of the Text Tokamak
NASA Astrophysics Data System (ADS)
Rhodes, Terry Lee
We studied the turbulence and fluctuation driven transport in the edge plasma of the TEXT tokamak using a Langmuir probe array. In this dissertation we present three separate experiments, each of which examines a particular aspect of the edge turbulence and transport. In the first experiment we compare the observed fluctuation levels to the scaling predictions of several turbulence theories. We found that the fluctuations and transport were not proportional to the density and temperature gradients. Thus, drift wave turbulence theories, which predict strong scalings with density gradients, do not describe the edge plasma turbulence. In the second experiment we identify low frequency modulations (<=q1KHz) in the edge density, potential and temperature to be associated with heat and density pulses (sawtooth oscillations) which originate from the central region of the tokamak. Concurrent with the edge sawtooth oscillations are significant increases in the density and potential fluctuation levels. As a result of these increases, the fluctuation driven particle flux and associated heat flux are increased as much as 60 and 100% respectively during the sawtooth. This result has direct implications on the current methods of determining the electron thermal diffusivity chi_ {e}. The effect of electron cyclotron heating (ECH) on the edge plasma was investigated in the third experiment. Increases in edge temperature, density and potential with simultaneous increases in the density and potential fluctuations were observed during ECH. These increased fluctuation levels resulted in a significant increase (20-50%) in the fluctuation driven particle flux. Comparison of these results to an equal input power, ohmic only discharge showed similar increases in the average density, temperature and potential. However, the density fluctuations did not increase as much with the additional ohmic heating (compared to ECH) resulting in a generally smaller comparative level of fluctuation
Mikhailovskii, A. B.; Novakovaskii, S. V.; Smolyakov, A. I.
1988-12-01
A theory is derived for the interaction of high-energy trapped particleswith ballooning modes in a tokamak with a high-..beta.. plasma. A dispersionrelation is derived to describe the ballooning modes in the presence ofsuch particles; the effects of the high plasma ..beta.. are taken into account.The stability boundary for ballooning modes with zero and finite frequenciesis studied. The effects of finite bananas on the stability of ballooningmodes with zero frequencies are determined.
NASA Astrophysics Data System (ADS)
Porkolab, M.; Lloyd, B.; Takase, Y.; Bonoli, P.; Fiore, C.; Gandy, R.; Granetz, R.; Griffin, D.; Gwinn, D.; Lipschultz, B.; Marmar, E.; McCool, S.; Pachtman, A.; Pappas, D.; Parker, R.; Pribyl, P.; Rice, J.; Terry, J.; Texter, S.; Watterson, R.; Wolfe, S.
1984-09-01
The effectiveness of plasma heating by electron Landau interaction in the lower hybrid range of frequencies in tokamak plasmas is demonstrated. Upon injection of 850 kW of rf power at a density of n―e~=1.4×1014 cm-3, an electron temperature increase of 1.0 keV and an ion temperature increase of 0.8 keV was achieved. These results are compared with transport and ray-tracing code predictions.
NASA Astrophysics Data System (ADS)
Jones, Morgin; Wadi, Hasina; Ali, Halima; Punjabi, Alkesh
2009-04-01
The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in divertor tokamaks can be any coordinates for which a transformation to (ψt,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψt is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct the symmetric quartic map such that the only parameter that determines magnetic geometry is the elongation of the separatrix surface. The poloidal flux inside the separatrix, the safety factor as a function of normalized minor radius, and the magnetic perturbation from the symplectic discretization are all held constant, and only the elongation is κ varied. The width of stochastic layer, the area, and the fractal dimension of the magnetic footprint and the average radial diffusion coefficient of magnetic field lines from the stochastic layer; and how these quantities scale with κ is calculated. The symmetric quartic map gives the correct scalings which are consistent with the scalings of coordinates with κ. The effects of m =1, n =±1 internal perturbation with the amplitude that is expected to occur in tokamaks are calculated by adding a term [H. Ali, A. Punjabi, A. H. Boozer, and T. Evans, Phys. Plasmas 11, 1908 (2004)] to the symmetric quartic map. In this case, the width of stochastic layer scales as 0.35 power of κ. The area of the footprint is roughly constant. The average radial diffusion coefficient of field lines near the X-point scales linearly with κ. The low mn perturbation changes the quasisymmetric structure of the footprint, and reorganizes it into a single, large scale, asymmetric structure. The symmetric quartic map is combined with the dipole map [A. Punjabi, H. Ali, and A. H. Boozer, Phys. Plasmas 10, 3992 (2003)] to calculate the effects of magnetic perturbation from a current carrying coil. The coil position and coil current coil are
Data Acquisition and Automation for Plasma Rotation Diagnostic in the TCABR Tokamak
NASA Astrophysics Data System (ADS)
Ronchi, G.; Severo, J. H. F.; de Sá, W. P.; Galvão, R. M. O.
2015-03-01
In this work we describe the implementation of a full modular system of data acquisition and processing for the plasma rotation diagnostic in the TCABR tokamak. The experimental setup uses a single monochromator and six photomultipliers (PMT), in which pair of PMTs measures the light at slightly different wavelengths. Thus, it can measure the time evolution of the Doppler shift of the impurities emission lines coming from three spatial positions (one for toroidal rotation and two for poloidal rotation). The data acquisition and preanalysis program were written with LabVIEW software and is capable of controlling the spectrometer wavelength, PMTs power supplies, data acquisition, and storage. All data are recorded in MDSplus trees that easily allow data visualization and post-processing analysis (both locally and remotely) via MATLAB, Python, Java and others programming languages. This system can run independently from other diagnostics and machine systems and can be integrated with the main tokamak control system by means of TCP/IP messages.
Particle control and plasma performance in the Lithium Tokamak eXperiment
Majeski, R.; Abrams, T.; Boyle, D.; Granstedt, E.; Hare, J.; Jacobson, C. M.; Kaita, R.; Kozub, T.; LeBlanc, B.; Lundberg, D. P.; Lucia, M.; Merino, E.; Schmitt, J.; Stotler, D.; Biewer, T. M.; Canik, J. M.; Gray, T. K.; Maingi, R.; McLean, A. G.; Kubota, S.; and others
2013-05-15
The Lithium Tokamak eXperiment is a small, low aspect ratio tokamak [Majeski et al., Nucl. Fusion 49, 055014 (2009)], which is fitted with a stainless steel-clad copper liner, conformal to the last closed flux surface. The liner can be heated to 350 °C. Several gas fueling systems, including supersonic gas injection and molecular cluster injection, have been studied and produce fueling efficiencies up to 35%. Discharges are strongly affected by wall conditioning. Discharges without lithium wall coatings are limited to plasma currents of order 10 kA, and discharge durations of order 5 ms. With solid lithium coatings discharge currents exceed 70 kA, and discharge durations exceed 30 ms. Heating the lithium wall coating, however, results in a prompt degradation of the discharge, at the melting point of lithium. These results suggest that the simplest approach to implementing liquid lithium walls in a tokamak—thin, evaporated, liquefied coatings of lithium—does not produce an adequately clean surface.
Nitrogen retention mechanisms in tokamaks with beryllium and tungsten plasma-facing surfaces
NASA Astrophysics Data System (ADS)
Oberkofler, M.; Meisl, G.; Hakola, A.; Drenik, A.; Alegre, D.; Brezinsek, S.; Craven, R.; Dittmar, T.; Keenan, T.; Romanelli, S. G.; Smith, R.; Douai, D.; Herrmann, A.; Krieger, K.; Kruezi, U.; Liang, G.; Linsmeier, Ch; Mozetic, M.; Rohde, V.; the ASDEX Upgrade Team; the EUROfusion MST1 Team; Contributors, JET
2016-02-01
Global gas balance experiments at ASDEX Upgrade (AUG) and JET have shown that a considerable fraction of nitrogen injected for radiative cooling is not recovered as N2 upon regeneration of the liquid helium cryo pump. The most probable loss channels are ion implantation into plasma-facing materials, co-deposition and ammonia formation. These three mechanisms are investigated in laboratory and tokamak experiments and by numerical simulations. Laboratory experiments have shown that implantation of nitrogen ions into beryllium and tungsten leads to the formation of surface nitrides, which may decompose under thermal loads. On beryllium the presence of nitrogen at the surface has been seen to reduce the sputtering yield. On tungsten surfaces it has been observed that the presence of nitrogen can increase hydrogen retention. The global nitrogen retention in AUG by implantation into the tungsten surfaces saturates. At JET the steady state nitrogen retention is increased by co-deposition with beryllium. The tokamak experiments are interpreted in detail by simulations of the global migration with WallDYN. Mass spectrometry of the exhaust gas of AUG and JET has revealed the conversion of nitrogen to ammonia at percent-levels. Conclusions are drawn on the potential implications of nitrogen seeding on the operation of a reactor in a deuterium-tritium mix.
NASA Astrophysics Data System (ADS)
Xu, X. Q.; Belli, E.; Bodi, K.; Candy, J.; Chang, C. S.; Cohen, R. H.; Colella, P.; Dimits, A. M.; Dorr, M. R.; Gao, Z.; Hittinger, J. A.; Ko, S.; Krasheninnikov, S.; McKee, G. R.; Nevins, W. M.; Rognlien, T. D.; Snyder, P. B.; Suh, J.; Umansky, M. V.
2009-06-01
We present edge gyrokinetic simulations of tokamak plasmas using the fully non-linear (full-f) continuum code TEMPEST. A non-linear Boltzmann model is used for the electrons. The electric field is obtained by solving the 2D gyrokinetic Poisson equation. We demonstrate the following. (1) High harmonic resonances (n > 2) significantly enhance geodesic-acoustic mode (GAM) damping at high q (tokamak safety factor), and are necessary to explain the damping observed in our TEMPEST q-scans and consistent with the experimental measurements of the scaling of the GAM amplitude with edge q95 in the absence of obvious evidence that there is a strong q-dependence of the turbulent drive and damping of the GAM. (2) The kinetic GAM exists in the edge for steep density and temperature gradients in the form of outgoing waves, its radial scale is set by the ion temperature profile, and ion temperature inhomogeneity is necessary for GAM radial propagation. (3) The development of the neoclassical electric field evolves through different phases of relaxation, including GAMs, their radial propagation and their long-time collisional decay. (4) Natural consequences of orbits in the pedestal and scrape-off layer region in divertor geometry are substantial non-Maxwellian ion distributions and parallel flow characteristics qualitatively like those observed in experiments.
Advances in multi-megawatt lower hybrid technology in support of steady-state tokamak operation
NASA Astrophysics Data System (ADS)
Delpech, L.; Achard, J.; Armitano, A.; Artaud, J. F.; Bae, Y. S.; Belo, J. H.; Berger-By, G.; Bouquey, F.; Cho, M. H.; Corbel, E.; Decker, J.; Do, H.; Dumont, R.; Ekedahl, A.; Garibaldi, P.; Goniche, M.; Guilhem, D.; Hillairet, J.; Hoang, G. T.; Kim, H. S.; Kim, J. H.; Kim, H.; Kwak, J. G.; Magne, R.; Mollard, P.; Na, Y. S.; Namkung, W.; Oh, Y. K.; Park, S.; Park, H.; Peysson, Y.; Poli, S.; Prou, M.; Samaille, F.; Yang, H. L.; The Tore Supra Team
2014-10-01
It has been demonstrated that lower hybrid current drive (LHCD) systems play a crucial role for steady-state tokamak operation, owing to their high current drive (CD) efficiency and hence their capability to reduce flux consumption. This paper describes the extensive technology programmes developed for the Tore Supra (France) and the KSTAR (Korea) tokamaks in order to bring continuous wave (CW) LHCD systems into operation. The Tore Supra LHCD generator at 3.7 GHz is fully CW compatible, with RF power PRF = 9.2 MW available at the generator to feed two actively water-cooled launchers. On Tore Supra, the most recent and novel passive active multijunction (PAM) launcher has sustained 2.7 MW (corresponding to its design value of 25 MW m-2 at the launcher mouth) for a 78 s flat-top discharge, with low reflected power even at large plasma-launcher gaps. The fully active multijunction (FAM) launcher has reached 3.8 MW of coupled power (24 MW m-2 at the launcher mouth) with the new TH2103C klystrons. By combining both the PAM and FAM launchers, 950 MJ of energy, using 5.2 MW of LHCD and 1 MW of ICRH (ion cyclotron resonance heating), was injected for 160 s in 2011. The 3.7 GHz CW LHCD system will be a key element within the W (for tungsten) environment in steady-state Tokamak (WEST) project, where the aim is to test ITER technologies for high heat flux components in relevant heat flux density and particle fluence conditions. On KSTAR, a 2 MW LHCD system operating at 5 GHz is under development. Recently the 5 GHz prototype klystron has reached 500 kW/600 s on a matched load, and studies are ongoing to design a PAM launcher. In addition to the studies of technology, a combination of ray-tracing and Fokker-Planck calculations have been performed to evaluate the driven current and the power deposition due to LH waves, and to optimize the N∥ spectrum for the future launcher design. Furthermore, an LHCD system at 5 GHz is being considered for a future upgrade of the ITER
Plasma interaction with tungsten samples in the COMPASS tokamak in ohmic ELMy H-modes
NASA Astrophysics Data System (ADS)
Dimitrova, M.; Weinzettl, V.; Matejicek, J.; Popov, Tsv; Marinov, S.; Costea, S.; Dejarnac, R.; Stöckel, J.; Havlicek, J.; Panek, R.
2016-03-01
This paper reports experimental results on plasma interaction with tungsten samples with or without pre-grown He fuzz. Under the experimental conditions, arcing was observed on the fuzzy tungsten samples, resulting in localized melting of the fuzz structure that did not extend into the bulk. The parallel power flux densities were obtained from the data measured by Langmuir probes embedded in the divertor tiles on the COMPASS tokamak. Measurements of the current-voltage probe characteristics were performed during ohmic ELMy H-modes reached in deuterium plasmas at a toroidal magnetic field BT = 1.15 T, plasma current Ip = 300 kA and line-averaged electron density ne = 5×1019 m-3. The data obtained between the ELMs were processed by the recently published first-derivative probe technique for precise determination of the plasma potential and the electron energy distribution function (EEDF). The spatial profile of the EEDF shows that at the high-field side it is Maxwellian with a temperature of 5 -- 10 eV. The electron temperatures and the ion-saturation current density obtained were used to evaluate the radial distribution of the parallel power flux density as being in the order of 0.05 -- 7 MW/m2.
Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak.
Soukhanovskii, V A; McLean, A G; Allen, S L
2014-11-01
New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and Te monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800-2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma Te, ne estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000-1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor Te monitoring aimed at divertor detachment real-time feedback control. PMID:25430325
Picosecond LIBS diagnostics for Tokamak in situ plasma facing materials chemical analysis
NASA Astrophysics Data System (ADS)
Morel, Vincent; Pérès, Bastien; Bultel, Arnaud; Hideur, Ammar; Grisolia, Christian
2016-02-01
First results are presented in relation with experimental and theoretical studies performed at the CORIA laboratory in the general framework of the determination of the chemical analysis of Tokamak plasma facing materials by laser-induced breakdown spectroscopy (LIBS) in picosecond regime. Experiments are performed on W in a specific chamber. This chamber is equipped with a UV-visible-near IR spectroscopic device. Boltzmann plots are derived for typical laser characteristics. We show that the initial excitation temperature is close to 12 000 K followed by a quasi steady value close to 8500 K. The ECHREM (Euler code for CHemically REactive Multicomponent laser-induced plasmas) code is developed to reproduce the laser-induced plasmas. This code is based on the implementation of a Collisional-Radiative model in which the different excited states are considered as full species. This state-to-state approach is relevant to theoretically assess the departure from excitation and chemical equilibrium. Tested on aluminum, the model shows that the plasma remains close to excitation equilibrium.
Near-infrared spectroscopy for divertor plasma diagnosis and control in DIII-D tokamak
Soukhanovskii, V. A. McLean, A. G.; Allen, S. L.
2014-11-15
New near infrared (NIR) spectroscopic measurements performed in the DIII-D tokamak divertor plasma suggest new viable diagnostic applications: divertor recycling and low-Z impurity flux measurements, a spectral survey for divertor Thomson scattering (DTS) diagnostic, and T{sub e} monitoring for divertor detachment control. A commercial 0.3 m spectrometer coupled to an imaging lens via optical fiber and a InGaAs 1024 pixel array detector enabled deuterium and impurity emission measurements in the range 800–2300 nm. The first full NIR survey identified D, He, B, Li, C, N, O, Ne lines and provided plasma T{sub e}, n{sub e} estimates from deuterium Paschen and Brackett series intensity and Stark line broadening analysis. The range 1.000–1.060 mm was surveyed in high-density and neon seeded divertor plasmas for spectral background emission studies for λ = 1.064 μm laser-based DTS development. The ratio of adjacent deuterium Paschen-α and Brackett Br9 lines in recombining divertor plasmas is studied for divertor T{sub e} monitoring aimed at divertor detachment real-time feedback control.
Cremaschini, Claudio; Tessarotto, Massimo
2011-11-15
A largely unsolved theoretical issue in controlled fusion research is the consistent kinetic treatment of slowly-time varying plasma states occurring in collisionless and magnetized axisymmetric plasmas. The phenomenology may include finite pressure anisotropies as well as strong toroidal and poloidal differential rotation, characteristic of Tokamak plasmas. Despite the fact that physical phenomena occurring in fusion plasmas depend fundamentally on the microscopic particle phase-space dynamics, their consistent kinetic treatment remains still essentially unchallenged to date. The goal of this paper is to address the problem within the framework of Vlasov-Maxwell description. The gyrokinetic treatment of charged particles dynamics is adopted for the construction of asymptotic solutions for the quasi-stationary species kinetic distribution functions. These are expressed in terms of the particle exact and adiabatic invariants. The theory relies on a perturbative approach, which permits to construct asymptotic analytical solutions of the Vlasov-Maxwell system. In this way, both diamagnetic and energy corrections are included consistently into the theory. In particular, by imposing suitable kinetic constraints, the existence of generalized bi-Maxwellian asymptotic kinetic equilibria is pointed out. The theory applies for toroidal rotation velocity of the order of the ion thermal speed. These solutions satisfy identically also the constraints imposed by the Maxwell equations, i.e., quasi-neutrality and Ampere's law. As a result, it is shown that, in the presence of nonuniform fluid and EM fields, these kinetic equilibria can sustain simultaneously toroidal differential rotation, quasi-stationary finite poloidal flows and temperature anisotropy.
The features of the global GAM in OH and ECRH plasmas in the T-10 tokamak
NASA Astrophysics Data System (ADS)
Melnikov, A. V.; Eliseev, L. G.; Perfilov, S. V.; Lysenko, S. E.; Shurygin, R. V.; Zenin, V. N.; Grashin, S. A.; Krupnik, L. I.; Kozachek, A. S.; Solomatin, R. Yu.; Elfimov, A. G.; Smolyakov, A. I.; Ufimtsev, M. V.; The HIBP Team
2015-06-01
Zonal flows and their high-frequency counterpart, the geodesic acoustic modes (GAMs) are considered as a possible mechanism of the plasma turbulence self-regulation. In the T-10 tokamak GAMs have been studied by the heavy ion beam probing and multipin Langmuir probes. The wide range of the regimes with Ohmic, on-axis and off-axis electron cyclotron resonance heating (ECRH) were studied (Bt = 1.5-2.4 T, Ip = 140-300 kA, \\bar{{n}}e = (0.6{--}6.0) × 1019 m-3 , PEC < 1.2 MW). It was shown that GAM has radially homogeneous structure and poloidal m = 0 for potential perturbations. The local theory predicts that fGAM ˜ \\sqrt {T/mi} /R , that means the frequency increases with the decrease of the minor radius. In contrast, the radial distribution of experimental frequency of the plasma potential and density oscillations, associated to GAM, is almost uniform over the whole plasma radius, suggesting the features of the nonlocal (global) eigenmodes. The GAM amplitude in the plasma potential also tends to be uniform along the radius. GAMs are more pronounced during ECRH, when the typical frequencies are seen in the narrow band from 22 to 27 kHz for the main peak and 25-30 kHz for the higher frequency satellite. GAM characteristics and the range of GAM existence are presented as functions of Te, density, magnetic field and PEC.
Mitarai, O; Xiao, C; McColl, D; Dreval, M; Hirose, A; Peng, M
2015-03-01
A plasma current up to 15 kA has been driven with outer ohmic heating (OH) coils in the STOR-M iron core tokamak. Even when the inner OH coil is disconnected, the outer OH coils alone can induce the plasma current as primary windings and initial breakdown are even easier in this coil layout. This result suggests a possibility to use an iron core in a spherical tokamak to start up the plasma current without a central solenoid. The effect of the iron core saturation on the extension of the discharge pulse length has been estimated for further experiments in the STOR-M tokamak. PMID:25832230
Pigarov, A Y; Krasheninnikov, S I; LaBombard, B; Rognlien, T D
2006-06-06
Large-Mach-number parallel plasma flows in the single-null SOL of different tokamaks are simulated with multi-fluid transport code UEDGE. The key role of poloidal asymmetry of cross-field plasma transport as the driving mechanism for such flows is discussed. The impact of ballooning-like diffusive and convective transport and plasma flows on divertor detachment, material migration, impurity flows, and erosion/deposition profiles is studied. The results on well-balanced double null plasma modeling that are indicative of strong asymmetry of cross-field transport are presented.
NASA Astrophysics Data System (ADS)
Mishra, Kishore; Idei, Hiroshi; Zushi, Hideki; Tashima, Saya; Banerjee, Santanu; Hasegawa, Makoto; Hanada, Kazuaki; Nakamura, Kazuo; Fujisawa, Akihide; Matsuoka, Keisuke; Nagashima, Yoshihiko; Kawasaki, S.; Higashijima, A.; Nakashima, H.
In spherical tokamak QUEST, plasma with high poloidal beta (βp) is obtained by injecting Electron Cyclotron Waves (ECW) into the Ohmic target plasma. With high βp, the plasma shape is transformed from an inboard limiter configuration to a natural divertor with the appearance of an poloidal null at the inboard side. By applying high vertical magnetic field (Bz), this high βp plasma is sustained. With suitable control of the equilibrium, effective current drive is observed in the later part of the discharge along with strong recharging of the Ohmic circuit and the Ip is sustained for > 1 s.
Darrow, Douglass S.; Ono, Masayuki
1990-03-06
A radial electric field of a desired magnitude and configuration is created throughout a substantial portion of the cross-section of the plasma of a tokamak. The radial electric field is created by injection of a unidirectional electron beam. The magnitude and configuration of the radial electric field may be controlled by the strength of the toroidal magnetic field of the tokamak.
Darrow, Douglass S.; Ono, Masayuki
1990-01-01
A radial electric field of a desired magnitude and configuration is created hroughout a substantial portion of the cross-section of the plasma of a tokamak. The radial electric field is created by injection of a unidirectional electron beam. The magnitude and configuration of the radial electric field may be controlled by the strength of the toroidal magnetic field of the tokamak.
Cui, Z. Q.; Chen, Z. J.; Xie, X. F.; Peng, X. Y.; Hu, Z. M.; Du, T. F.; Ge, L. J.; Zhang, X.; Yuan, X.; Fan, T. S.; Chen, J. X.; Li, X. Q. E-mail: guohuizhang@pku.edu.cn; Zhang, G. H. E-mail: guohuizhang@pku.edu.cn; Xia, Z. W.; Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N.
2014-11-15
The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.
Cui, Z Q; Chen, Z J; Xie, X F; Peng, X Y; Hu, Z M; Du, T F; Ge, L J; Zhang, X; Yuan, X; Xia, Z W; Hu, L Q; Zhong, G Q; Lin, S Y; Wan, B N; Fan, T S; Chen, J X; Li, X Q; Zhang, G H
2014-11-01
The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G. PMID:25430242
NASA Astrophysics Data System (ADS)
Cui, Z. Q.; Chen, Z. J.; Xie, X. F.; Peng, X. Y.; Hu, Z. M.; Du, T. F.; Ge, L. J.; Zhang, X.; Yuan, X.; Xia, Z. W.; Hu, L. Q.; Zhong, G. Q.; Lin, S. Y.; Wan, B. N.; Fan, T. S.; Chen, J. X.; Li, X. Q.; Zhang, G. H.
2014-11-01
The novel neutron spectrometer TOFED (Time of Flight Enhanced Diagnostics), comprising 90 individual photomultiplier tubes coupled with 85 plastic scintillation detectors through light guides, has been constructed and installed at Experimental Advanced Superconducting Tokamak. A dedicated magnetic shielding system has been constructed for TOFED, and is designed to guarantee the normal operation of photomultiplier tubes in the stray magnetic field leaking from the tokamak device. Experimental measurements and numerical simulations carried out employing the finite element method are combined to optimize the design of the magnetic shielding system. The system allows detectors to work properly in an external magnetic field of 200 G.
Advanced limiter test (ALT-1) in the TEXTOR tokamak: concept and experimental design
Conn, R.W.; Grontz, S.P.; Prinja, A.K.; Gauster, W.B.; Malinowski, H.E.; Pontau, A.E.; Blewer, R.S.; Whitley, J.B.; Dippel, K.H.; Fuchs, G.
1983-01-01
The concept and experimental design of a pump-limiter for the TEXTOR tokamak is described. The module is constructed of stainless steel with a compound curvature head designed to limit the maximum heat flux to 300 W/cm/sup 2/. The head is made of TiC-coated graphite containing a variable-aperture slot to admit plasma to a deflector plate for ballistic pumping action. The assembly is actively pumped using Zr-Al getters with an estimated hydrogen pumping speed of 3 x 10/sup 4/ 1/s. The aspect ratio of the pump duct and the length of the plasma channel are both variable to permit study of plasma plugging, ballistic scattering, and enhanced gas-conduction effects. The module can be moved radially by 10 cm to permit its operation either as the primary or secondary limiter. Major diagnostics include Langmuir and solid state probes, bolometers, infrared thermography, thermocouples, ion gauges, manometers, and a gas mass analyzer.
NASA Astrophysics Data System (ADS)
Dendy, R. O.; McClements, K. G.
2015-04-01
Ion cyclotron emission (ICE) was the first collective radiative instability, driven by confined fusion-born ions, observed from deuterium-tritium plasmas in JET and TFTR. ICE comprises strongly suprathermal emission, which has spectral peaks at multiple ion cyclotron harmonic frequencies as evaluated at the outer mid-plane edge of tokamak plasmas. The measured intensity of ICE spectral peaks scaled linearly with measured fusion reactivity in JET. In other large tokamak plasmas, ICE is currently used as an indicator of fast ions physics. The excitation mechanism for ICE is the magnetoacoustic cyclotron instability (MCI); in the case of JET and TFTR, the MCI is driven by a set of centrally born trapped fusion products, lying just inside the trapped-passing boundary in velocity space, whose drift orbits make large radial excursions to the outer mid-plane edge. Diagnostic exploitation of ICE in future experiments therefore rests in part on deep understanding of the MCI, and recent advances in computational plasma physics have led to substantial recent progress, reviewed here. Particle-in-cell simulations of the MCI, with fully kinetic ions and electrons, were reported in 2013, using plasma parameters for JET ICE observations. The hybrid approximation for plasma simulations, where ions are treated as particles and electrons as a neutralising massless fluid, was then applied and reported in 2014. These simulations extend previous studies deep into the nonlinear regime of the MCI, and corroborate predictions by linear analytical theory, thereby strengthening further the link to ICE measurements. ICE is a potential diagnostic for confined alpha-particles in ITER, where measurements of ICE could yield information on energetic ion behaviour supplementing that obtainable from other diagnostics. In addition, it may be possible to use ICE to study fast ion redistribution and loss due to MHD activity in ITER.
NASA Astrophysics Data System (ADS)
Wang, G. Q.; Ma, J.; Weiland, J.; Zang, Q.
2013-10-01
We have made the first drift wave study of particle transport in the Experimental Advanced Superconducting Tokamak (Wan et al., Nucl. Fusion 49, 104011 (2009)). The results reveal that collisions make the particle flux more inward in the high collisionality regime. This can be traced back to effects that are quadratic in the collision frequency. The particle pinch is due to electron trapping which is not very efficient in the high collisionality regime so the approach to equilibrium is slow. We have included also the electron temperature gradient (ETG) mode to give the right electron temperature gradient, since the Trapped Electron Mode (TE mode) is weak in this regime. However, at the ETG mode number ions are Boltzmann distributed so the ETG mode does not give particle transport.
Conceptual design of a fast-ion D-alpha diagnostic on experimental advanced superconducting tokamak.
Huang, J; Heidbrink, W W; Wan, B; von Hellermann, M G; Zhu, Y; Gao, W; Wu, C; Li, Y; Fu, J; Lyu, B; Yu, Y; Shi, Y; Ye, M; Hu, L; Hu, C
2014-11-01
To investigate the fast ion behavior, a fast ion D-alpha (FIDA) diagnostic system has been planned and is presently under development on Experimental Advanced Superconducting Tokamak. The greatest challenges for the design of a FIDA diagnostic are its extremely low intensity levels, which are usually significantly below the continuum radiation level and several orders of magnitude below the bulk-ion thermal charge-exchange feature. Moreover, an overlaying Motional Stark Effect (MSE) feature in exactly the same wavelength range can interfere. The simulation of spectra code is used here to guide the design and evaluate the diagnostic performance. The details for the parameters of design and hardware are presented. PMID:25430314
Gyrokinetic study of electromagnetic effects on toroidal momentum transport in tokamak plasmas
Hein, T.; Angioni, C.; Fable, E.; Candy, J.; Peeters, A. G.
2011-07-15
The effect of a finite {beta}{sub e} = 8{pi}n{sub e}T{sub e}/B{sup 2} on the turbulent transport of toroidal momentum in tokamak plasmas is discussed. From an analytical gyrokinetic model as well as local linear gyrokinetic simulations, it is shown that the modification of the parallel mode structure due to the nonadiabatic response of passing electrons, which changes the parallel wave vector k{sub ||} with increasing {beta}{sub e}, leads to a decrease in size of both the diagonal momentum transport as well as the Coriolis pinch under ion temperature gradient turbulence conditions, while for trapped electron modes, practically no modification is found. The decrease is particularly strong close to the onset of the kinetic ballooning modes. There, the Coriolis pinch even reverses its direction.
NASA Astrophysics Data System (ADS)
Zhong, W. L.; Shen, Y.; Zou, X. L.; Gao, J. M.; Shi, Z. B.; Dong, J. Q.; Duan, X. R.; Xu, M.; Cui, Z. Y.; Li, Y. G.; Ji, X. Q.; Yu, D. L.; Cheng, J.; Xiao, G. L.; Jiang, M.; Yang, Z. C.; Zhang, B. Y.; Shi, P. W.; Liu, Z. T.; Song, X. M.; Ding, X. T.; Liu, Yong; HL-2A Team
2016-07-01
The impact of impurity ions on a pedestal has been investigated in the HL-2A Tokamak, at the Southwestern Institute of Physics, Chengdu, China. Experimental results have clearly shown that during the H -mode phase, an electromagnetic turbulence was excited in the edge plasma region, where the impurity ions exhibited a peaked profile. It has been found that double impurity critical gradients are responsible for triggering the turbulence. Strong stiffness of the impurity profile has been observed during cyclic transitions between the I -phase and H -mode regime. The results suggest that the underlying physics of the self-regulated edge impurity profile offers the possibility for an active control of the pedestal dynamics via pedestal turbulence.
Alfvén acoustic channel for ion energy in high-beta tokamak plasmas.
Bierwage, Andreas; Aiba, Nobuyuki; Shinohara, Kouji
2015-01-01
When the plasma beta (ratio of thermal to magnetic pressure) in the core of a tokamak is raised to values of several percent, as required for a thermonuclear fusion reactor, continuous spectra of long-wavelength slow magnetosonic waves enter the frequency band occupied by continuous spectra of shear Alfvén waves. It is found that these two branches can couple strongly, so that Alfvén modes that are resonantly driven by suprathermal ions transfer some of their energy to sound waves. Since sound waves are heavily damped by thermal ion Landau resonances, these results reveal a new energy channel that contributes to the damping of Alfvénic instabilities and the noncollisional heating of bulk ions, with potentially important consequences for confinement and fusion performance. PMID:25615474
Alfvén Acoustic Channel for Ion Energy in High-Beta Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Bierwage, Andreas; Aiba, Nobuyuki; Shinohara, Kouji
2015-01-01
When the plasma beta (ratio of thermal to magnetic pressure) in the core of a tokamak is raised to values of several percent, as required for a thermonuclear fusion reactor, continuous spectra of long-wavelength slow magnetosonic waves enter the frequency band occupied by continuous spectra of shear Alfvén waves. It is found that these two branches can couple strongly, so that Alfvén modes that are resonantly driven by suprathermal ions transfer some of their energy to sound waves. Since sound waves are heavily damped by thermal ion Landau resonances, these results reveal a new energy channel that contributes to the damping of Alfvénic instabilities and the noncollisional heating of bulk ions, with potentially important consequences for confinement and fusion performance.
M3D-K simulations of sawteeth and energetic particle transport in tokamak plasmas
NASA Astrophysics Data System (ADS)
Shen, Wei; Fu, G. Y.; Sheng, Zheng-Mao; Breslau, J. A.; Wang, Feng
2014-09-01
Nonlinear simulations of sawteeth and related energetic particle transport are carried out using the kinetic/magnetohydrodynamic (MHD) hybrid code M3D-K. MHD simulations show repeated sawtooth cycles for a model tokamak equilibrium. Furthermore, test particle simulations are carried out to study the energetic particle transport due to a sawtooth crash. The results show that energetic particles are redistributed radially in the plasma core, depending on pitch angle and energy. For trapped particles, the redistribution occurs for particle energy below a critical value in agreement with existing theories. For co-passing particles, the redistribution is strong with little dependence on particle energy. In contrast, the redistribution level of counter-passing particles decreases with increasing particle energy.
M3D-K Simulations of Sawteeth and Energetic Particle Transport in Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Shen, Wei; Fu, Guoyong; Sheng, Zhengmao; Breslau, Joshua; Wang, Feng
2013-10-01
Nonlinear simulations of Sawteeth and energetic particle transport are carried out using the kinetic/MHD hybrid code M3D-K. MHD simulations show repeated sawtooth cycles due to a resistive (1,1) internal kink mode for a model tokamak equilibrium. Furthermore, test particle simulations are carried out to study the energetic particle transport due to a sawtooth crash. The results show that energetic particles are redistributed radially in plasma core depending on pitch angle and energy. For trapped particles, the redistribution occurs for particle energy below a critical value in agreement with previous theory. For co-passing particles, the redistribution is strong with little dependence on particle energy. In contrast, the redistribution level of counter-passing particles decreases as particle energy becomes large.
M3D-K simulations of sawteeth and energetic particle transport in tokamak plasmas
Shen, Wei; Sheng, Zheng-Mao; Fu, G. Y.; Breslau, J. A.; Wang, Feng
2014-09-15
Nonlinear simulations of sawteeth and related energetic particle transport are carried out using the kinetic/magnetohydrodynamic (MHD) hybrid code M3D-K. MHD simulations show repeated sawtooth cycles for a model tokamak equilibrium. Furthermore, test particle simulations are carried out to study the energetic particle transport due to a sawtooth crash. The results show that energetic particles are redistributed radially in the plasma core, depending on pitch angle and energy. For trapped particles, the redistribution occurs for particle energy below a critical value in agreement with existing theories. For co-passing particles, the redistribution is strong with little dependence on particle energy. In contrast, the redistribution level of counter-passing particles decreases with increasing particle energy.
Zhong, W L; Shen, Y; Zou, X L; Gao, J M; Shi, Z B; Dong, J Q; Duan, X R; Xu, M; Cui, Z Y; Li, Y G; Ji, X Q; Yu, D L; Cheng, J; Xiao, G L; Jiang, M; Yang, Z C; Zhang, B Y; Shi, P W; Liu, Z T; Song, X M; Ding, X T; Liu, Yong
2016-07-22
The impact of impurity ions on a pedestal has been investigated in the HL-2A Tokamak, at the Southwestern Institute of Physics, Chengdu, China. Experimental results have clearly shown that during the H-mode phase, an electromagnetic turbulence was excited in the edge plasma region, where the impurity ions exhibited a peaked profile. It has been found that double impurity critical gradients are responsible for triggering the turbulence. Strong stiffness of the impurity profile has been observed during cyclic transitions between the I-phase and H-mode regime. The results suggest that the underlying physics of the self-regulated edge impurity profile offers the possibility for an active control of the pedestal dynamics via pedestal turbulence. PMID:27494476
Toroidal rotation of multiple species of ions in tokamak plasma driven by lower-hybrid-waves
Zuo Yang; Wang Shaojie; Pan Chengkang
2012-10-15
A numerical simulation is carried out to investigate the toroidal rotation of multiple species of ions and the radial electric field in a tokamak plasma driven by the lower-hybrid-wave (LHW). The theoretical model is based on the neoclassical transport theory associated with the anomalous transport model. Three species of ions (primary ion and two species of impurity ions) are taken into consideration. The predicted toroidal velocity of the trace impurities during the LHW injection agrees reasonably well with the experimental observation. It is shown that the toroidal rotation velocities of the trace impurity ions and the primary ions are close, therefore the trace impurity ions are representative of the primary ions in the toroidal rotation driven by the LHW.
NASA Technical Reports Server (NTRS)
Bhatia, A. K.; Feldman, U.; Doschek, G. A.
1980-01-01
The paper presents calculations of electron impact collision strengths and spontaneous radiative decay rates for titanium ions of the LiI through FI isoelectronic sequences for transitions between levels of the 2S(2)2p(k), 2s2p(k+1), and 2p(k+2) configurations. From these atomic data, excitation-rate coefficients are calculated along with level populations for these three configurations. The calculations of level populations include the effects of proton excitation, and are carried out at electron temperatures and densities typical of tokamak plasmas. Wavelengths of forbidden and intersystem lines are given, and a synthetic spectrum is presented for a typical temperature and density.
Effects of toroidal field ripple on suprathermal ions in tokamak plasmas
Goldston, R.J.; Towner, H.H.
1980-02-01
Analytic calculations of three important effects of toroidal field ripple on suprathermal ions in tokamak plasmas are presented. In the first process, collisional ripple-trapping, beam ions become trapped in local magnetic wells near their banana tips due to pitch-angle scattering as they traverse the ripple on barely unripple-trapped orbits. In the second process, collisionless ripple-trapping, near-perpendicular untrapped ions are captured (again near a banana tip) due to their finite orbits, which carry them out into regions of higher ripple. In the third process, banana-drift diffusion, fast-ion banana orbits fail to close precisely, due to a ripple-induced variable lingering period near the banana tips. These three mechanisms lead to substantial radial transport of banana-trapped, neutral-beam-injected ions when the quantity ..cap alpha..* identical with epsilon/sin theta/Nqdelta is of order unity or smaller.
Compact retarding field energy analyzer for the tokamak ISTTOK boundary plasma
Nedzelskiy, I. S.; Silva, C.; Figueiredo, H.; Fernandes, H.; Varandas, C. A. F.
2006-10-15
The retarding field energy analyzer (RFEA) remains the more reliable diagnostic to measure the ion temperature in the boundary plasmas of magnetic fusion devices. A compact, simple design and inexpensive RFEA has been developed for investigations on the tokamak ISTTOK. It consists of a stainless steel pinhole (with a diameter of 0.6 mm), three fine nickel grids with a separation of 1 mm, and a collector, all insulated by mica. All the components are placed inside a boron nitride housing with dimensions of 14x14x23 mm{sup 3}. The RFEA has been tested in both ion and electron modes. The conditions of the RFEA operation are discussed, and preliminary measurements of the ion and electron temperature profiles presented.
A resistive magnetodynamics analysis of sawtooth driven tearing modes in tokamak plasmas
NASA Astrophysics Data System (ADS)
Guo, Wenping; Wang, Jiaqi; Liu, Dongjian; Wang, Xiaogang
2016-06-01
In this paper, a resistive magnetohydrodynamics model is applied to study the effect of sawtooth driven on classical/neoclassical tearing modes in tokamak plasmas. In a model of forced reconnection, the sawtooth is considered as a boundary disturbance for m >1 modes and causes the islands growth of m/n = 2/1 and 3/2 modes through toroidal coupling. Theoretical and numerical analyses show that the linear growth of the modes is driven by precursors of the sawtooth through the linear mode coupling, while differential rotation has great effect on both the linear and the nonlinear development of the modes. It is believed that the tearing mode can be suppressed by control of the sawtooth by radio frequency heating or current drive.
Noninductive plasma generation and current drive in the Globus-M spherical tokamak
D'yachenko, V. V.; Gusev, V. K.; Larionov, M. M.; Mel'nik, A. D.; Novokhatskii, A. N.; Petrov, Yu. V.; Rozhdestvenskii, V. V.; Sakharov, N. V.; Stepanov, A. Yu.; Khitrov, S. A.; Khromov, N. A.; Chernyshev, F. V.; Shevelev, A. E.; Shcherbinin, O. N.; Bender, S. E.; Kavin, A. A.; Lobanov, K. M.
2013-03-15
Experimental results on the generation and maintenance of the toroidal current in the Globus-M spherical tokamak by using waves in the lower hybrid frequency range without applying an inductive vortex electric field are presented. For this purpose, the original ridge guide antennas forming a field distribution similar to that produced by multiwaveguide grills were used. The high-frequency field (900 MHz) was used for both plasma generation and current drive. The magnitude of the generated current reached 21 kA, and its direction depended on the direction of the vertical magnetic field. Analysis of the experimental results indicates that the major fraction of the current is carried by the suprathermal electron beam.
Movable multi-probes for plasma boundary measurement in sino-united spherical tokamak
Chai, Song Wang, Wenhao; Tan, Yi; Gao, Zhe
2014-11-15
A novel movable multi-probes is developed to get local magnetic and electrostatic profiles on Sino-UNIted Spherical Tokamak (SUNIST). This multi-probes combines a four-tips Langmuir probe, a magnetic coil, and a retarding field energy analyzer (RFEA). It can be used to simultaneously measure the poloidal magnetic field B{sub p}, electric field E{sub r}, electron temperature T{sub e}, electron density n{sub e}, and ion temperature T{sub i}. Its small overall size (20 × 20 × 38 mm{sup 3}) enables the movable multi-probes to measure the magnetic and electrostatic profiles in high spatial resolution, with negligible impact to plasma in SUNIST. This paper presents the design of the movable multi-probes, in particular, details of RFEA for reliable ion energy measurements. Preliminary experimental results of the movable multi-probes are given as well.
Advances in Plasma-Filled Microwave Sources
NASA Astrophysics Data System (ADS)
Goebel, Dan M.
1998-11-01
Significant improvements in the performance of high power microwave tubes have been achieved in recent years by the introduction of plasma into the beam- coupling structures of the devices. Plasma has been credited with increasing the maximum electron beam current, frequency bandwidth, electrical efficiency and reducing or eliminating the need for guiding magnetic fields in microwave sources. These advances are critically important for the development of high power, frequency agile microwave systems where size and weight are important. Conversely, plasma has been blamed for causing noise, instabilities, power variations and pulse-length limitations in microwave tubes for many years. Recent experimental and theoretical studies have demonstrated that introducing the right amount of plasma in a controlled manner can be beneficial in the areas described above. Enhanced beam propagation at lower magnetic fields and higher beam current levels due to the space-charge neutralization by plasma can be realized provided that the neutralization fraction is fairly stable and maintained near a value of one for the duration of the desired pulse length. The generation of hybrid waves in plasma-filled slow-wave structures (SWS) operating near cutoff has resulted in an increased electric field on axis and improved coupling to solid beams in both helix and coupled-cavity SWS, and wider coupling-aperture pass-bands and frequency bandwidth in coupled-cavity devices. In the event of excess plasma generation in these TWTs or BWOs, the device structures rapidly approach cutoff or breakdown and the beam forms instabilities, which degrades the output power level and pulse length. Recent experimental and theoretical advances in this field including plasma implementation techniques in the gun and circuit will be presented, and the benefits and limitations of plasma filling of microwave sources will be shown and discussed.