Resistive wall mode stabilization by plasma rotation in advanced tokamaks
NASA Astrophysics Data System (ADS)
Eriksson, G.
1996-03-01
By combining previous results of Betti and Freidberg [Phys. Rev. Lett. 74, 2949 (1995)] and Eriksson [Phys. Plasmas 2, 3095 (1995)], a fully analytical description is obtained for the stabilizing effect of toroidal plasma rotation in a large aspect ratio tokamak surrounded by a resistive wall. As in advanced tokamak configurations with a large fraction of bootstrap current, it is assumed that the current gradient near the plasma edge is large. This assumption enables an analytical analysis of external kink modes with low poloidal mode numbers. An expression is obtained, showing explicitly how the window of stable wall distances depends on the current profile.
Saturated internal instabilities in advanced-tokamak plasmas
NASA Astrophysics Data System (ADS)
Hua, M.-D.; Chapman, I. T.; Pinches, S. D.; Hastie, R. J.; MAST Team
2010-06-01
"Advanced tokamak" (AT) scenarios were developed with the aim of reaching steady-state operation in future potential tokamak fusion power plants. AT scenarios exhibit non-monotonic to flat safety factor profiles (q, a measure of the magnetic field line pitch), with the minimum q (qmin) slightly above an integer value (qs). However, it has been predicted that these q profiles are unstable to ideal magnetohydrodynamic instabilities as qmin approaches qs. These ideal instabilities, observed and diagnosed as such for the first time in MAST plasmas with AT-like q profiles, have far-reaching consequences like confinement degradation, flattening of the toroidal core rotation or enhanced fast ion losses. These observations motivate the stability analysis of advanced-tokamak plasmas, with a view to provide guidance for stability thresholds in AT scenarios. Additionally, the measured rotation damping is compared to the self-consistently calculated predictions from neoclassical toroidal viscosity theory.
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.
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.
NASA Astrophysics Data System (ADS)
Nam, Y. U.; Chung, J.
2010-10-01
A 280 GHz single-channel horizontal millimeter-wave interferometer system has been installed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR) device. This system has a triangular beam path that does not pass through the plasma axis due to geometrical constraints in the superconducting tokamak. The term line density on KSTAR has a different meaning from the line density of other tokamaks. To estimate the peak density and the mean density from the measured line density, information on the position of the plasma is needed. The information has been calculated from tangentially viewed visible images using the toroidal symmetry of the plasma. Interface definition language routines have been developed for this purpose. The calculated plasma position data correspond well to calculation results from magnetic analysis. With the position data and an estimated plasma profile, the peak density and the mean density have been obtained from the line density. From these results, changes of plasma density themselves can be separated from effects of the plasma movements, so they can give valuable information on the plasma status.
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.
Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma
NASA Astrophysics Data System (ADS)
Xu, Liqing; Zhang, Jizong; Chen, Kaiyun; Hu, Liqun; Li, Erzhong; Lin, Shiyao; Shi, Tonghui; Duan, Yanmin; Zhu, Yubao
2015-12-01
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.
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.
An advanced plasma control system for the DIII-D tokamak
Ferron, J.R.; Kellman, A.; McKee, E.; Osborne, T.; Petrach, P.; Taylor, T.S.; Wight, J. ); Lazarus, E. )
1991-11-01
An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Development of this system is expected to lead to control system technology appropriate for use on future tokamaks such as ITER and BPX. The digital system will allow for increased precision in shape control through real time adjustment of the control algorithm to changes in the shape and discharge parameters such as {beta}{sub p}, {ell}{sub i} and scrape-off layer current. The system will be used for research on real time optimization of discharge performance for disruption avoidance, current and pressure profile control, optimization of rf antenna loading, or feedback on heat deposition patterns through divertor strike point position control, for example. Shape control with this system is based on linearization near a target shape of the controlled parameters as a function of the magnetic diagnostic signals. This digital system is unique in that it is designed to have the speed necessary to control the unstable vertical motion of highly elongated tokamak discharges such as those produced in DIII-D and planned for BPX and ITER. a 40 MHz Intel i860 processor is interfaced to up to 112 channels of analog input signals. The commands to the poloidal field coils can be updated at 80 {mu}s intervals for the control of vertical position with a delay between sampling of the analog signal and update of the command of less than 80 {mu}s.
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.
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.
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.
NASA Astrophysics Data System (ADS)
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.; East Team
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.
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.
Plasma models for real-time control of advanced tokamak scenarios
NASA Astrophysics Data System (ADS)
Moreau, D.; Mazon, D.; Walker, M. L.; Ferron, J. R.; Burrell, K. H.; Flanagan, S. M.; Gohil, P.; Groebner, R. J.; Hyatt, A. W.; La Haye, R. J.; Lohr, J.; Turco, F.; Schuster, E.; Ou, Y.; Xu, C.; Takase, Y.; Sakamoto, Y.; Ide, S.; Suzuki, T.; ITPA-IOS Group members; experts
2011-06-01
An integrated plasma profile control strategy, ARTAEMIS, is being developed for extrapolating present-day advanced tokamak (AT) scenarios to steady-state operation. The approach is based on semi-empirical modelling and was initially explored on JET (Moreau et al 2008 Nucl. Fusion 48 106001). This paper deals with the general applicability of this strategy for simultaneous magnetic and kinetic control on various tokamaks. The determination of the device-specific, control-oriented models that are needed to compute optimal controller matrices for a given operation scenario is discussed. The methodology is generic and can be applied to any device, with different sets of heating and current drive actuators, controlled variables and profiles. The system identification algorithms take advantage of the large ratio between the magnetic and thermal diffusion time scales and have been recently applied to both JT-60U and DIII-D data. On JT-60U, an existing series of high bootstrap current (~70%), 0.9 MA non-inductive AT discharges was used. The actuators consisted of four groups of neutral beam injectors aimed at perpendicular injection (on-axis and off-axis), and co-current tangential injection (also on-axis and off-axis). On DIII-D, dedicated system identification experiments were carried out in the loop voltage (Vext) control mode (as opposed to current control) to avoid feedback in the response data from the primary circuit. The reference plasma state was that of a 0.9 MA AT scenario which had been optimized to combine non-inductive current fractions near unity with 3.5 < βN < 3.9, bootstrap current fractions larger than 65% and H98(y,2) = 1.5. Actuators other than Vext were co-current, counter-current and balanced neutral beam injection, and electron cyclotron current drive. Power and loop voltage modulations resulted in dynamic variations of the plasma current between 0.7 and 1.2 MA. It is concluded that the response of essential plasma parameter profiles to specific
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.
Advanced commercial tokamak study
Thomson, S.L.; Dabiri, A.E.; Keeton, D.C.; Brown, T.G.; Bussell, G.T.
1985-12-01
Advanced commercial tokamak studies were performed by the Fusion Engineering Design Center (FEDC) as a participant in the Tokamak Power Systems Studies (TPSS) project coordinated by the Office of Fusion Energy. The FEDC studies addressed the issues of tokamak reactor cost, size, and complexity. A scoping study model was developed to determine the effect of beta on tokamak economics, and it was found that a competitive cost of electricity could be achieved at a beta of 10 to 15%. The implications of operating at a beta of up to 25% were also addressed. It was found that the economics of fusion, like those of fission, improve as unit size increases. However, small units were found to be competitive as elements of a multiplex plant, provided that unit cost and maintenance time reductions are realized for the small units. The modular tokamak configuration combined several new approaches to develop a less complex and lower cost reactor. The modular design combines the toroidal field coil with the reactor structure, locates the primary vacuum boundary at the reactor cell wall, and uses a vertical assembly and maintenance approach. 12 refs., 19 figs.
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.
Advanced tokamak operating modes in TPX and ITER
Nevins, W.M.
1994-12-31
A program is described to develop the advanced tokamak physics required for an economic steady-state fusion reactor on existing (short-pulse) tokamak experiments; to extend these operating modes to long-pulse on TPX; and finally to demonstrate them in a long-pulse D-T plasma on ITER.
Transport Equations In Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Callen, J. D.
2009-11-01
Tokamak plasma transport equations are usually obtained by flux surface averaging the collisional Braginskii equations. However, tokamak plasmas are not in collisional regimes. Also, ad hoc terms are added for: neoclassical effects on the parallel Ohm's law (trapped particle effects on resistivity, bootstrap current); fluctuation-induced transport; heating, current-drive and flow sources and sinks; small B field non-axisymmetries; magnetic field transients etc. A set of self-consistent second order in gyroradius fluid-moment-based transport equations for nearly axisymmetric tokamak plasmas has been developed recently using a kinetic-based framework. The derivation uses neoclassical-based parallel viscous force closures, and includes all the effects noted above. Plasma processes on successive time scales (and constraints they impose) are considered sequentially: compressional Alfv'en waves (Grad-Shafranov equilibrium, ion radial force balance); sound waves (pressure constant along field lines, incompressible flows within a flux surface); and ion collisions (damping of poloidal flow). Radial particle fluxes are driven by the many second order in gyroradius toroidal angular torques on the plasma fluid: 7 ambipolar collision-based ones (classical, neoclassical, etc.) and 8 non-ambipolar ones (fluctuation-induced, polarization flows from toroidal rotation transients etc.). The plasma toroidal rotation equation [1] results from setting to zero the net radial current induced by the non-ambipolar fluxes. The radial particle flux consists of the collision-based intrinsically ambipolar fluxes plus the non-ambipolar fluxes evaluated at the ambipolarity-enforcing toroidal plasma rotation (radial electric field). The energy transport equations do not involve an ambipolar constraint and hence are more directly obtained. The resultant transport equations will be presented and contrasted with the usual ones. [4pt] [1] J.D. Callen, A.J. Cole, C.C. Hegna, ``Toroidal Rotation In
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.
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.
Toroidal Flow in Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Callen, J. D.; Cole, A. J.; Hegna, C. C.
2007-11-01
Many effects influence toroidal flow evolution in tokamak plasmas. Momentum sources and radial diffusion due to axisymmetric neoclassical, paleoclassical and anomalous transport are usually considered. In addition, the toroidal flow can be affected by field errors. Small, non-axisymmetric field errors arise from coil irregularities, active control coils and collective plasma magnetic distortions (e.g., NTMs, RWMs). Resonant field errors cause localized electromagnetic torques near rational surfaces in the plasma, which can lock the plasma to the wall leading to magnetic islands and reduced confinement or disruptions. Their penetration into the plasma is limited by flow-shielding effects; but they can be amplified by the plasma response at high beta. Non-resonant field errors cause magnetic pumping and radial banana drifts, and lead to toroidal flow damping over the entire plasma. Many of these processes can also produce momentum pinch and intrinsic flow effects. This poster will seek to present a coherent picture of all these effects and suggest ways they could be tested and distinguished experimentally.
Predictive transport simulations of real-time profile control in JET advanced tokamak plasmas
NASA Astrophysics Data System (ADS)
Tala, T.; Laborde, L.; Mazon, D.; Moreau, D.; Corrigan, G.; Crisanti, F.; Garbet, X.; Heading, D.; Joffrin, E.; Litaudon, X.; Parail, V.; Salmi, A.; EFDA-JET workprogramme, contributors to the
2005-09-01
Predictive, time-dependent transport simulations with a semi-empirical plasma model have been used in closed-loop simulations to control the q-profile and the strength and location of the internal transport barrier (ITB). Five transport equations (Te, Ti, q, ne, vΦ) are solved, and the power levels of lower hybrid current drive, NBI and ICRH are calculated in a feedback loop determined by the feedback controller matrix. The real-time control (RTC) technique and algorithms used in the transport simulations are identical to those implemented and used in JET experiments (Laborde L. et al 2005 Plasma Phys. Control. Fusion 47 155 and Moreau D. et al 2003 Nucl. Fusion 43 870). The closed-loop simulations with RTC demonstrate that varieties of q-profiles and pressure profiles in the ITB can be achieved and controlled simultaneously. The simulations also showed that with the same RTC technique as used in JET experiments, it is possible to sustain the q-profiles and pressure profiles close to their set-point profiles for longer than the current diffusion time. In addition, the importance of being able to handle the multiple time scales to control the location and strength of the ITB is pointed out. Several future improvements and perspectives of the RTC scheme are presented.
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.
Fusion Plasma Theory: Task 3, Auxiliary radiofrequency heating of tokamaks
Scharer, J.E.
1992-01-01
The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.
Toi, K.; Ogawa, K.; Isobe, M.; Osakabe, M.; Spong, Donald A; Todo, Yasushi
2011-01-01
Comprehensive understanding of energetic-ion-driven global instabilities such as Alfven eigenmodes (AEs) and their impact on energetic ions and bulk plasma is crucially important for tokamak and stellarator/helical plasmas and in the future for deuterium-tritium (DT) burning plasma experiments. Various types of global modes and their associated enhanced energetic ion transport are commonly observed in toroidal plasmas. Toroidicity-induced AEs and ellipticity-induced AEs, whose gaps are generated through poloidal mode coupling, are observed in both tokamak and stellarator/helical plasmas. Global AEs and reversed shear AEs, where toroidal couplings are not as dominant were also observed in those plasmas. Helicity induced AEs that exist only in 3D plasmas are observed in the large helical device (LHD) and Wendelstein 7 Advanced Stellarator plasmas. In addition, the geodesic acoustic mode that comes from plasma compressibility is destabilized by energetic ions in both tokamak and LHD plasmas. Nonlinear interaction of these modes and their influence on the confinement of the bulk plasma as well as energetic ions are observed in both plasmas. In this paper, the similarities and differences in these instabilities and their consequences for tokamak and stellarator/helical plasmas are summarized through comparison with the data sets obtained in LHD. In particular, this paper focuses on the differences caused by the rotational transform profile and the 2D or 3D geometrical structure of the plasma equilibrium. Important issues left for future study are listed.
NASA Astrophysics Data System (ADS)
Toi, K.; Ogawa, K.; Isobe, M.; Osakabe, M.; Spong, D. A.; Todo, Y.
2011-02-01
Comprehensive understanding of energetic-ion-driven global instabilities such as Alfvén eigenmodes (AEs) and their impact on energetic ions and bulk plasma is crucially important for tokamak and stellarator/helical plasmas and in the future for deuterium-tritium (DT) burning plasma experiments. Various types of global modes and their associated enhanced energetic ion transport are commonly observed in toroidal plasmas. Toroidicity-induced AEs and ellipticity-induced AEs, whose gaps are generated through poloidal mode coupling, are observed in both tokamak and stellarator/helical plasmas. Global AEs and reversed shear AEs, where toroidal couplings are not as dominant were also observed in those plasmas. Helicity induced AEs that exist only in 3D plasmas are observed in the large helical device (LHD) and Wendelstein 7 Advanced Stellarator plasmas. In addition, the geodesic acoustic mode that comes from plasma compressibility is destabilized by energetic ions in both tokamak and LHD plasmas. Nonlinear interaction of these modes and their influence on the confinement of the bulk plasma as well as energetic ions are observed in both plasmas. In this paper, the similarities and differences in these instabilities and their consequences for tokamak and stellarator/helical plasmas are summarized through comparison with the data sets obtained in LHD. In particular, this paper focuses on the differences caused by the rotational transform profile and the 2D or 3D geometrical structure of the plasma equilibrium. Important issues left for future study are listed.
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.
Nondiffusive plasma transport at tokamak edge
NASA Astrophysics Data System (ADS)
Krasheninnikov, S. I.
2000-10-01
Recent findings show that cross field edge plasma transport at tokamak edge does not necessarily obey a simple diffusive law [1], the only type of a transport model applied so far in the macroscopic modeling of edge plasma transport. Cross field edge transport is more likely due to plasma filamentation with a ballistic motion of the filaments towards the first wall. Moreover, it so fast that plasma recycles on the main chamber first wall rather than to flow into divertor as conventional picture of edge plasma fluxes suggests. Crudely speaking particle recycling wise diverted tokamak operates in a limiter regime due to fast anomalous non-diffusive cross field plasma transport. Obviously that this newly found feature of edge plasma anomalous transport can significantly alter a design of any future reactor relevant tokamaks. Here we present a simple model describing the motion of the filaments in the scrape off layer and discuss it implications for experimental observations. [1] M. Umansky, S. I. Krasheninnikov, B. LaBombard, B. Lipschultz, and J. L. Terry, Phys. Plasmas 6 (1999) 2791; M. Umansky, S. I. Krasheninnikov, B. LaBombard and J. L. Terry, Phys. Plasmas 5 (1998) 3373.
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.
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.
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.
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.
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.
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.
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.
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.
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
Advanced tokamak scenario developments for the next step
NASA Astrophysics Data System (ADS)
Joffrin, E.
2007-12-01
The objective of advanced tokamak scenario research is to provide a candidate plasma scenario for continuous operation in a fusion power plant. The optimization of the self-generated non-inductive current by the bootstrap mechanism up to a level of 50% and above using high plasma pressure and improved confinement are the necessary conditions to achieve this goal. The two main candidate scenarios for continuous operation, the steady state scenario and long duration (up to 3000 s) high neutron fluency scenario (the hybrid scenario), both face physics challenges in terms of confinement, stability, power exhaust and plasma control. Resistive wall modes and Alfvénic fast ion driven instabilities are the main limitations for operating the steady state scenario at high pressure and low magnetic shear. In addition, this scenario demands a high degree of control over the plasma current and pressure profile and the steady state heat load on in-vessel plasma facing components. Understanding the confinement properties of hybrid scenario is still an outstanding issue as well as its modelling for ITER in particular with regard to the H-mode pedestal parameters. This scenario will also require active current profile control, although, less demanding than for the steady state scenario. To operate advanced tokamak scenario, broad current and pressure profile control appears as a necessary requirement on ITER actuators, in addition to the tools required for instability control such as error field coils or electron cyclotron current drive.
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.
Mathematical modeling plasma transport in tokamaks
Quiang, Ji
1997-01-01
In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10^{20}/m^{3} with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.
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.
Viscosity in the edge of tokamak plasmas
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.
An efficient transport solver for tokamak plasmas
Park, Jin Myung; Murakami, Masanori; St. John, H. E.; ...
2017-01-03
A simple approach to efficiently solve a coupled set of 1-D diffusion-type transport equations with a stiff transport model for tokamak plasmas is presented based on the 4th order accurate Interpolated Differential Operator scheme along with a nonlinear iteration method derived from a root-finding algorithm. Here, numerical tests using the Trapped Gyro-Landau-Fluid model show that the presented high order method provides an accurate transport solution using a small number of grid points with robust nonlinear convergence.
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.
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.
A simulation study of a controlled tokamak plasma
NASA Astrophysics Data System (ADS)
Fujii, N.; Niwa, Y.
1980-03-01
A tokamak circuit theory, including results of numerical simulation studies, is applied to a control system synthesized for a Joule heated tokamak plasma. The treatment is similar to that of Ogata and Ninomiya (1979) except that in this case a quadrupole field coil current is considered coexisting with image induced on a vacuum chamber.
Forced magnetic reconnection in Tokamak plasmas
NASA Astrophysics Data System (ADS)
Cole, Andrew Joseph
This dissertation addresses two related problems in the study of forced magnetic reconnection in Tokamak plasmas. First, a recent controversy concerning a model forced magnetic reconnection problem, the Taylor problem, has been resolved. The criticisms of Ishizawa and Tokuda [21] concerning the original analysis of Hahm and Kulsrud [17] are shown to be unwarranted, both analytically and numerically. Second, one possible reason for the discrepancy between recent experimental [29] and previous theoretical [13] scaling of the critical error-field penetration threshold with device parameters is addressed. The theory in question is entirely based on a single-fluid MHD (magnetohydrodynamical) treatment of the plasma. As is well-known, high temperature plasmas are far better modeled using the drift-MHD ordering.[18] Hence we develop a drift-MHD theory of error-field penetration. Although two new drift-MHD plasma response regimes are identified, the overall threshold scaling with device parameters is not altogether different from that predicted by single-fluid MHD.
New DIII-D tokamak plasma control system
Campbell, G.L.; Ferron, J.R.; McKee, E.; Nerem, A.; Smith, T; Greenfield, C.M.; Pinsker, R.I. ); Lazarus, E.A. JET Joint Undertaking, Abingdon, Oxon )
1992-09-01
A state-of-the-art plasma control system has been constructed for use on the DIII-D tokamak to provide high speed real time data acquisition and feedback control of DIII-D plasma parameters. This new system has increased the precision to which discharge shape and position parameters can be maintained and has provided the means to rapidly change from one plasma configuration to another. The capability to control the plasma total energy and the ICRF antenna loading resistance has been demonstrated. The speed and accuracy of this digital system will allow control of the current drive and heating systems in order to regulate the current and pressure profiles and diverter power deposition in the DIII-D machine. Use of this system will allow the machine and power supplies to be better protected from undesirable operating regimes. The advanced control system is also suitable for control algorithm development for future machines in these areas and others such as disruption avoidance. The DIII-D tokamak facility is operated for the US Department of Energy by General Atomics Company (GA) in San Diego, California. The DIII-D experimental program will increase emphasis on rf heating and current drive in the near future and is installing a cryopumped divertor ring during the fall of 1992. To improve the flexibility of this machine for these experiments, the new shape control system was implemented. The new advanced plasma control system has enhanced the capabilities of the DIII-D machine and provides a data acquisition and control platform that promises to be useful far beyond its original charter.
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.
Nusselt number scaling in tokamak plasma turbulence
Takeda, K.; Benkadda, S.; Hamaguchi, S.; Wakatani, M.
2005-05-15
Anomalous heat transport caused by ion temperature gradient (ITG) driven turbulence in tokamak plasmas is evaluated from numerical simulations of the two-dimensional (2D) partial-differential equations of the ITG model and of a reduced 1D version derived from a quasilinear approximation. In the strongly turbulent state, intermittent bursts of thermal transport are observed in both cases. In the strongly turbulent regime, the reduced model as well as the direct numerical simulation show that the Nusselt number Nu (normalized heat flux) scales with the normalized ion pressure gradient K{sub i} as Nu{proportional_to}K{sub i}{sup 1/3}. Since the Rayleigh number for ITG turbulence is proportional to K{sub i}, the Nusselt number scaling for ITG turbulence is thus similar to the classical thermal transport scaling for Rayleigh-Benard convections in neutral fluids.
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.
Summary discussion: An integrated advanced tokamak reactor
Sauthoff, N.R.
1994-12-31
The tokamak concept improvement workshop addressed a wide range of issues involved in the development of a more attractive tokamak. The agenda for the workshop progressed from a general discussion of the long-range energy context (with the objective being the identification of a set of criteria and ``figures of merit`` for measuring the attractiveness of a tokamak concept) to particular opportunities for the improvement of the tokamak concept. The discussions concluded with a compilation of research program elements leading to an improved tokamak concept.
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.
Small angle slot divertor concept for long pulse advanced tokamaks
NASA Astrophysics Data System (ADS)
Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.
2017-04-01
SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.
Filterscope diagnostic system on the Experimental Advanced Superconducting Tokamak (EAST).
Xu, Z; Wu, Z W; Gao, W; Chen, Y J; Wu, C R; Zhang, L; Huang, J; Chang, J F; Yao, X J; Gao, W; Zhang, P F; Jin, Z; Hou, Y M; Guo, H Y
2016-11-01
A filterscope diagnostic system has been mounted to observe the line emission and visible bremsstrahlung emission from plasma on the experimental advanced superconducting tokamak during the 2014 campaign. By this diagnostic system, multiple wavelengths including Dα (656.1 nm), Dγ (433.9 nm), He ii (468.5 nm), Li i (670.8 nm), Li ii (548.3 nm), C iii (465.0 nm), O ii (441.5 nm), Mo i (386.4 nm), W i (400.9 nm), and visible bremsstrahlung radiation (538.0 nm) are monitored with corresponding wavelength filters. All these multi-channel signals are digitized at up to 200 kHz simultaneously. This diagnostic plays a crucial role in studying edge localized modes and H-mode plasmas, due to the high temporal resolution and spatial resolution that have been designed into it.
NASA Astrophysics Data System (ADS)
Seo, Seong-Heon; Park, Jinhyung; Wi, H. M.; Lee, W. R.; Kim, H. S.; Lee, T. G.; Kim, Y. S.; Kang, Jin-Seob; Bog, M. G.; Yokota, Y.; Mase, A.
2013-08-01
Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 1019 m-3 when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.
Seo, Seong-Heon; Park, Jinhyung; Wi, H M; Lee, W R; Kim, H S; Lee, T G; Kim, Y S; Kang, Jin-Seob; Bog, M G; Yokota, Y; Mase, A
2013-08-01
Frequency modulation reflectometer has been developed to measure the plasma density profile of the Korea Superconducting Tokamak Advanced Research tokamak. Three reflectometers are operating in extraordinary polarization mode in the frequency range of Q band (33.6-54 GHz), V band (48-72 GHz), and W band (72-108 GHz) to measure the density up to 7 × 10(19) m(-3) when the toroidal magnetic field is 2 T on axis. The antenna is installed inside of the vacuum vessel. A new vacuum window is developed by using 50 μm thick mica film and 0.1 mm thick gold gasket. The filter bank of low pass filter, notch filter, and Faraday isolator is used to reject the electron cyclotron heating high power at attenuation of 60 dB. The full frequency band is swept in 20 μs. The mixer output is directly digitized with sampling rate of 100 MSamples/s. The phase is obtained by using wavelet transform. The whole hardware and software system is described in detail and the measured density profile is presented as a result.
Sensitivity of predictive tokamak plasma transport simulations
Redd, A.J.; Kritz, A.H.; Bateman, G.; Kinsey, J.E.
1997-06-01
The sensitivity of our time-dependent simulations of low confinement (L-mode) discharges to variations in initial profiles and time-dependent boundary conditions has been explored. These time-dependent tokamak plasma simulations were performed using a theory-based Multi-mode transport model that includes ion temperature gradient (ITG) and trapped electron modes (TEM), kinetic and resistive ballooning modes and neoclassical modes. The density and temperature profiles predicted in our simulations of L-mode discharges are found to be robust, even with significant variations in the initial or boundary conditions. Although transport associated with a single mode can be strongly affected by local changes in plasma parameters resulting from changes in the boundary conditions, the total transport remains largely unchanged because of compensation by other transport modes. The sensitivity of the predicted temperature and density profiles to a variation in the Multi-mode model is also examined. When the Dominguez-Waltz theory of transport driven by ITG and TEM modes is replaced in the Multi-mode model by the Weiland description, we find that the predictions of the Weiland model more closely match the experimental data. {copyright} {ital 1997 American Institute of Physics.}
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.
Diagnosing transient plasma status: from solar atmosphere to tokamak divertor
NASA Astrophysics Data System (ADS)
Giunta, A. S.; Henderson, S.; O'Mullane, M.; Harrison, J.; Doyle, J. G.; Summers, H. P.
2016-09-01
This work strongly exploits the interdisciplinary links between astrophysical (such as the solar upper atmosphere) and laboratory plasmas (such as tokamak devices) by sharing the development of a common modelling for time-dependent ionisation. This is applied to the interpretation of solar flare data observed by the UVSP (Ultraviolet Spectrometer and Polarimeter), on-board the Solar Maximum Mission and the IRIS (Interface Region Imaging Spectrograph), and also to data from B2-SOLPS (Scrape Off Layer Plasma Simulations) for MAST (Mega Ampère Spherical Tokamak) Super-X divertor upgrade. The derived atomic data, calculated in the framework of the ADAS (Atomic Data and Analysis Structure) project, allow equivalent prediction in non-stationary transport regimes and transients of both the solar atmosphere and tokamak divertors, except that the tokamak evolution is about one thousand times faster.
Status of neutron diagnostics on the experimental advanced superconducting tokamak
NASA Astrophysics Data System (ADS)
Zhong, G. Q.; Hu, L. Q.; Pu, N.; Zhou, R. J.; Xiao, M.; Cao, H. R.; Zhu, Y. B.; Li, K.; Fan, T. S.; Peng, X. Y.; Du, T. F.; Ge, L. J.; Huang, J.; Xu, G. S.; Wan, B. N.
2016-11-01
Neutron diagnostics have become a significant means to study energetic particles in high power auxiliary heating plasmas on the Experimental Advanced Superconducting Tokamak (EAST). Several kinds of neutron diagnostic systems have been implemented for time-resolved measurements of D-D neutron flux, fluctuation, emission profile, and spectrum. All detectors have been calibrated in laboratory, and in situ calibration using 252Cf neutron source in EAST is in preparation. A new technology of digitized pulse signal processing is adopted in a wide dynamic range neutron flux monitor, compact recoil proton spectrometer, and time of flight spectrometer. Improvements will be made continuously to the system to achieve better adaptation to the EAST's harsh γ-ray and electro-magnetic radiation environment.
Status of neutron diagnostics on the experimental advanced superconducting tokamak.
Zhong, G Q; Hu, L Q; Pu, N; Zhou, R J; Xiao, M; Cao, H R; Zhu, Y B; Li, K; Fan, T S; Peng, X Y; Du, T F; Ge, L J; Huang, J; Xu, G S; Wan, B N
2016-11-01
Neutron diagnostics have become a significant means to study energetic particles in high power auxiliary heating plasmas on the Experimental Advanced Superconducting Tokamak (EAST). Several kinds of neutron diagnostic systems have been implemented for time-resolved measurements of D-D neutron flux, fluctuation, emission profile, and spectrum. All detectors have been calibrated in laboratory, and in situ calibration using (252)Cf neutron source in EAST is in preparation. A new technology of digitized pulse signal processing is adopted in a wide dynamic range neutron flux monitor, compact recoil proton spectrometer, and time of flight spectrometer. Improvements will be made continuously to the system to achieve better adaptation to the EAST's harsh γ-ray and electro-magnetic radiation environment.
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.
The ARIES Advanced and Conservative Tokamak Power Plant Study
Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...
2015-12-22
Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦtotal N of 5.75, an H98 of 1.65, anmore » n/nGr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦtotalN of 2.5, an H₉₈ of 1.25, an n/nGr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less
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 (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.
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.
Diamagnetic loop measurement in Korea Superconducting Tokamak Advanced Research machine.
Bak, J G; Lee, S G; Kim, H S
2011-06-01
Diamagnetic loop (DL), which consists of two poloidal loops inside the vacuum vessel, is used to measure the diamagnetic flux during a plasma discharge in the Korea Superconducting Tokamak Advanced Research (KSTAR) machine. The vacuum fluxes in the DL signal can be compensated up to 0.1 mWb by using the coefficients, which are obtained from experimental investigations, in the vacuum flux measurements during vacuum shots under same operational conditions of magnetic coils for plasma experiment in the KSTAR machine. The maximum error in the diamagnetic flux measurement due to the errors of the coefficients was estimated as ∼0.22 mWb. From the diamagnetic flux measurements for the ohmically heated circular plasmas in the KSTAR machine, the stored energy agrees well with the estimated kinetic energy within the discrepancy of 25%. When the electron cyclotron heating, the neutral beam injection, and the ion cyclotron resonance heating are added to the ohmically heated limiter plasmas, the additional heating effects can be clearly observed from the increase of the stored energy evaluated in the DL measurement.
Diamagnetic loop measurement in Korea Superconducting Tokamak Advanced Research machine
Bak, J. G.; Lee, S. G.; Kim, H. S.
2011-06-15
Diamagnetic loop (DL), which consists of two poloidal loops inside the vacuum vessel, is used to measure the diamagnetic flux during a plasma discharge in the Korea Superconducting Tokamak Advanced Research (KSTAR) machine. The vacuum fluxes in the DL signal can be compensated up to 0.1 mWb by using the coefficients, which are obtained from experimental investigations, in the vacuum flux measurements during vacuum shots under same operational conditions of magnetic coils for plasma experiment in the KSTAR machine. The maximum error in the diamagnetic flux measurement due to the errors of the coefficients was estimated as {approx}0.22 mWb. From the diamagnetic flux measurements for the ohmically heated circular plasmas in the KSTAR machine, the stored energy agrees well with the estimated kinetic energy within the discrepancy of 25%. When the electron cyclotron heating, the neutral beam injection, and the ion cyclotron resonance heating are added to the ohmically heated limiter plasmas, the additional heating effects can be clearly observed from the increase of the stored energy evaluated in the DL measurement.
System studies for quasi-steady-state advanced physics tokamak
Reid, R.L.; Peng, Y.K.M.
1983-11-01
Parametric studies were conducted using the Fusion Engineering Design Center (FEDC) Tokamak Systems Code to investigate the impact of veriation in physics parameters and technology limits on the performance and cost of a low q/sub psi/, high beta, quasi-steady-state tokamak for the purpose of fusion engineering experimentation. The features and characteristics chosen from each study were embodied into a single Advanced Physics Tokamak design for which a self-consistent set of parameters was generated and a value of capital cost was estimated.
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.
Stochastic modeling of plasma mode forecasting in tokamak
NASA Astrophysics Data System (ADS)
Saadat, Sh.; Salem, M.; Ghoranneviss, M.; Khorshid, P.
2012-04-01
The structure of magnetohydrodynamic (MHD) modes has always been an interesting study in tokamaks. The mode number of tokamak plasma is the most important parameter, which plays a vital role in MHD instabilities. If it could be predicted, then the time of exerting external fields, such as feedback fields and Resonance Helical Field, could be obtained. Autoregressive Integrated Moving Average (ARIMA) and Seasonal Autoregressive Integrated Moving Average are useful models to predict stochastic processes. In this paper, we suggest using ARIMA model to forecast mode number. The ARIMA model shows correct mode number (m = 4) about 0.5 ms in IR-T1 tokamak and equations of Mirnov coil fluctuations are obtained. It is found that the recursive estimates of the ARIMA model parameters change as the plasma mode changes. A discriminator function has been proposed to determine plasma mode based on the recursive estimates of model parameters.
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.
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.
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.
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.
NASA Astrophysics Data System (ADS)
Kukushkin, A. B.; Rantsev-Kartinov, V. A.
1999-02-01
The method of multilevel dynamical contrasting is applied to analyzing available data from tokamak plasmas. The results illustrate a possibility of extending the concept of the plasma percolating networks in dense Z pinches (and other inertially confined plasmas) to the case of magnetically confined plasmas. This extension suggests a necessity to append the conventional picture of the nonfilamentary plasma (which is nearly a fluid described by conventional magnetohydrodynamics) with a "network" component which is formed by the strongest long-living filaments of electric current and penetrate the "fluid" component. Signs of networking are found in visible light and soft x-ray images, and magnetic probing data. A diagnostic algorithm is formulated for identifying the role of plasma networking in observed phenomena of nonlocal (non-diffusive) heat transport in a tokamak.
Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989
None
2016-07-12
From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.
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.
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.
Neural net prediction of tokamak plasma disruptions
NASA Astrophysics Data System (ADS)
Hernandez, J. V.; Lin, Z.; Horton, W.; Vannucci, A.; McCool, S. C.
1994-10-01
The computation based on neural net algorithms in predicting minor and major disruptions in TEXT tokamak discharges has been performed. Future values of the fluctuating magnetic signal are predicted based on L past values of the magnetic fluctuation signal, measured by a single Mirnov coil. The time step used (= 0.04ms) corresponds to the experimental data sampling rate. Two kinds of approaches are adopted for the task, the contiguous future prediction and the multi-timescale prediction. Results are shown for comparison. Both networks are trained through the back-propagation algorithm with inertial terms. The degree of this success indicates that the magnetic fluctuations associated with tokamak disruptions may be characterized by a relatively low-dimensional dynamical system.
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 new low drift integrator system for the Experiment Advanced Superconductor Tokamak.
Liu, D M; Wan, B N; Wang, Y; Wu, Y C; Shen, B; Ji, Z S; Luo, J R
2009-05-01
A new type of the integrator system with the low drift characteristic has been developed to accommodate the long pulse plasma discharges on Experiment Advanced Superconductor Tokamak (EAST). The integrator system is composed of the Ethernet control module and the integral module which includes one integrator circuit, followed by two isolation circuits and two program-controlled amplifier circuits. It compensates automatically integration drift and is applied in real-time control. The performance test and the experimental results in plasma discharges show that the developed integrator system can meet the requirements of plasma control on the accuracy and noise level of the integrator in long pulse discharges.
Plasma shaping effects on tokamak scrape-off layer turbulence
NASA Astrophysics Data System (ADS)
Riva, Fabio; Lanti, Emmanuel; Jolliet, Sébastien; Ricci, Paolo
2017-03-01
The impact of plasma shaping on tokamak scrape-off layer (SOL) turbulence is investigated. The drift-reduced Braginskii equations are written for arbitrary magnetic geometries, and an analytical equilibrium model is used to introduce the dependence of turbulence equations on tokamak inverse aspect ratio (ε ), Shafranov’s shift (Δ), elongation (κ), and triangularity (δ). A linear study of plasma shaping effects on the growth rate of resistive ballooning modes (RBMs) and resistive drift waves (RDWs) reveals that RBMs are strongly stabilized by elongation and negative triangularity, while RDWs are only slightly stabilized in non-circular magnetic geometries. Assuming that the linear instabilities saturate due to nonlinear local flattening of the plasma gradient, the equilibrium gradient pressure length {L}p=-{p}e/{{\
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
Shape reconstruction of merging spherical tokamak plasma in UTST device
NASA Astrophysics Data System (ADS)
Ushiki, Tomohiko; Itagaki, Masafumi; Inomoto, Michiaki
2016-10-01
Spherical tokamak (ST) merging method is one of the ST start-up methods which heats the plasma through magnetic reconnection. In the present study reconstruction of eddy current profile and plasma shape was performed during spherical tokamak merging only using external sensor signals by the Cauchy condition surface (CCS) method. CCS method have been implemented for JT-60 (QST), QUEST (Kyushu University), KSTAR (NFRI), RELAX (KIT), and LHD (Nifs). In this method, CCS was assumed inside each plasmas, where both flux function and its normal derivative are unknown. Effect of plasma current was replaced by the boundary condition of CCS, assuming vacuum field everywhere. Also, the nodal points for the boundary integrals of eddy current density were set using quadratic elements in order to express the complicated vacuum vessel shape. Reconstructed profiles of the eddy current and the magnetic flux were well coincided with the reference in each phase of merging process. Magnetic sensor installation plan for UTST was determined from these calculation results. This work was supported by the JSPS A3 Foresight Program ``Innovative Tokamak Plasma Startup and Current Drive in Spherical Torus''.
NASA Astrophysics Data System (ADS)
Humphreys, D. A.; Ferron, J. R.; Leuer, J. A.; Walker, M. L.; Welander, A. S.
2003-10-01
Linear, perturbed equilibrium plasma response models can accurately represent the experimental response of tokamak plasmas to applied fields [A. Coutlis, et al., Nucl. Fusion 39, 663 (1999)]. However, agreement between experiment and model is much better when average flux over the plasma, rather than at each fluid element, is conserved [P. Vyas, et al., Nucl. Fusion 38, 1043 (1998)]. The close experimental agreement of average flux-conserving models is consistent with approximating field penetration effects produced by finite plasma resistivity, particularly in the edge region. We report on the development of nonrigid linear plasma response models which include finite local plasma resistivity in order to more accurately represent the dynamic response due to this field penetration. Such response models are expected to be important for designing profile control algorithms in advanced tokamaks. Accounting for finite plasma resistivity is also important in designing multivariable integrated controllers which must simultaneously regulate plasma shape and plasma current. Consequences of including resisitivity will be illustrated and comparisons with DIII-D experimental plasma responses will be made.
Understanding L–H transition in tokamak fusion plasmas
NASA Astrophysics Data System (ADS)
Guosheng, XU; Xingquan, WU
2017-03-01
This paper reviews the current state of understanding of the L–H transition phenomenon in tokamak plasmas with a focus on two central issues: (a) the mechanism for turbulence quick suppression at the L–H transition; (b) the mechanism for subsequent generation of sheared flow. We briefly review recent advances in the understanding of the fast suppression of edge turbulence across the L–H transition. We uncover a comprehensive physical picture of the L–H transition by piecing together a number of recent experimental observations and insights obtained from 1D and 2D simulation models. Different roles played by diamagnetic mean flow, neoclassical-driven mean flow, turbulence-driven mean flow, and turbulence-driven zonal flows are discussed and clarified. It is found that the L–H transition occurs spontaneously mediated by a shift in the radial wavenumber spectrum of edge turbulence, which provides a critical evidence for the theory of turbulence quench by the flow shear. Remaining questions and some key directions for future investigations are proposed. This work was supported by National Magnetic Confinement Fusion Science Program of China under Contracts No. 2015GB101000, No. 2013GB106000, and No. 2013GB107000 and National Natural Science Foundation of China under Contracts No. 11575235 and No. 11422546.
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.
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.
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
Public Data Set: Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup
Hinson, Edward T.; Barr, Jayson L.; Bongard, Michael W.; Burke, Marcus G.; Fonck, Raymond J.; Perry, Justin M.
2016-05-31
This data set contains openly-documented, machine readable digital research data corresponding to figures published in E.T. Hinson et al., 'Impedance of an Intense Plasma-Cathode Electron Source for Tokamak Plasma Startup,' Physics of Plasmas 23, 052515 (2016).
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.
Momentum injection in tokamak plasmas and transitions to reduced transport.
Parra, F I; Barnes, M; Highcock, E G; Schekochihin, A A; Cowley, S C
2011-03-18
The effect of momentum injection on the temperature gradient in tokamak plasmas is studied. A plausible scenario for transitions to reduced transport regimes is proposed. The transition happens when there is sufficient momentum input so that the velocity shear can suppress or reduce the turbulence. However, it is possible to drive too much velocity shear and rekindle the turbulent transport. The optimal level of momentum injection is determined. The reduction in transport is maximized in the regions of low or zero magnetic shear.
Multi-field plasma sandpile model in tokamaks and applications
NASA Astrophysics Data System (ADS)
Peng, X. D.; Xu, J. Q.
2016-08-01
A multi-field sandpile model of tokamak plasmas is formulated for the first time to simulate the dynamic process with interaction between avalanche events on the fast/micro time-scale and diffusive transports on the slow/macro time-scale. The main characteristics of the model are that both particle and energy avalanches of sand grains are taken into account simultaneously. New redistribution rules of a sand-relaxing process are defined according to the transport properties of special turbulence which allows the uphill particle transport. Applying the model, we first simulate the steady-state plasma profile self-sustained by drift wave turbulences in the Ohmic discharge of a tokamak. A scaling law as f = a q0 b + c for the relation of both center-density n ( 0 ) and electron (ion) temperatures T e ( 0 ) ( T i ( 0 ) ) with the center-safety-factor q 0 is found. Then interesting work about the nonlocal transport phenomenon observed in tokamak experiments proceeds. It is found that the core electron temperature increases rapidly in response to the edge cold pulse and inversely it decreases in response to the edge heat pulse. The results show that the nonlocal response of core electron temperature depending on the amplitudes of background plasma density and temperature is more remarkable in a range of gas injection rate. Analyses indicate that the avalanche transport caused by plasma drift instabilities with thresholds is a possible physical mechanism for the nonlocal transport in tokamaks. It is believed that the model is capable of being applied to more extensive questions occurring in the transport field.
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.
Heavy Neutral Beam Probe for Edge Plasma Analysis in Tokamaks
Castracane, J.
2001-01-04
The Heavy Neutral Beam Probe (HNBP) developed initially with DOE funding under the Small Business Innovation Research (SBIR) program was installed on the Tokamak de Varennes (TdeV) at the CCFM. This diagnostic was designed to perform fundamental measurements of edge plasma properties. The hardware was capable of measuring electron density and potential profiles with high spatial and temporal resolution. Fluctuation spectra for these parameters were obtained with HNBP for transport studies.
Observation of Pedestal Plasma Turbulence on EAST Tokamak
NASA Astrophysics Data System (ADS)
Gao, Xiang; Zhang, Tao; Han, Xiang; Zhang, Shoubiao; Wang, Yumin; Liu, Zixi; Yang, Yao; Liu, Shaocheng; Shi, Nan; Ling, Bili; Li, Jiangang; The EAST Team
2013-08-01
Pedestal plasma turbulence was experimentally studied by microwave reflectometry on EAST tokamak. The characteristics of edge pedestal turbulence during dithering L-H transition, ELM-free H-mode phase and inter-ELM phase have recently been studied on EAST. An edge spatial structure of density fluctuation and its dithering temporal evolution is observed for the first time on the EAST tokamak during the L-H transition phase. A coherent mode usually appears during the ELM-free phase prior to the first ELM on EAST tokamak. The mode frequency gradually decreases as the pedestal evolves. Analysis shows that the coherent mode is in the pedestal region inside the separatrix. In plasma with type-III ELMs, a precursor mode before ELM is usually observed. The frequency of the precursor was initially about 150 kHz and gradually decreased till the next ELM. The mode amplitude increases or shows saturation before ELM. In the plasma with compound ELMs composed of high and low frequency ELMs, the precursor was also observed before the high frequency ELM while the harmonic oscillations with frequencies of 20 kHz, 40 kHz and 60 kHz appear before the low frequency ELM.
Robust control of long-pulse, high performance plasmas in KSTAR tokamak
NASA Astrophysics Data System (ADS)
Jeon, Youngmu; Hahn, S. H.; Han, H. S.; Woo, M. H.; Joung, M.; Kim, Jayhyun; Bae, Y. S.; Kim, H.-S.; Yoon, S. W.; Oh, Y. K.; Na, Y. S.; Eidietis, N. W.; Walker, M. L.; Lanctot, M. J.; Hyatt, A. W.; Mueller, D. A.; Kstar Team
2016-10-01
The goal of KSTAR is to achieve and demonstrate high performance, steady state tokamak operations in long pulse up to 300 s. In recent years, we made significant progresses on plasma control and performance for this advanced tokamak (AT) operation. First of all, the plasma equilibrium magnetic control has been substantially improved by applying fully decoupled multi-input-multi-output (MIMO) isoflux shape controllers [1]. The MIMO shape controllers were designed using a newly developed design method by taking the plasma equilibrium response into account self-consistently. More than eight shape control variables including plasma currents are controlled independently on each other with high accuracy (less than 1cm error on average) and with wide variations of plasma shape. By virtue of this robust control, various long pulse H-mode discharges have been operated up to 60 s, which was the maximum pulse length allowable in current KSTAR system. Also, plasma performance has been improved accordingly. A fully non-inductive H-mode operation [1] was achieved for the first time in KSTAR, through the so-called `high betap' operation with betap 3.0. In addition, various experimental attempts for advanced scenario development have been conducted such as the `hybrid' [2] and `high li' scenarios[3].
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.
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.
Spectra of heliumlike krypton from Tokamak Fusion Test Reactor plasmas
Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M. ); Beiersdorfer, P.; Osterheld, A. ); Smith, A. ); Fraenkel, B. )
1993-08-16
Experiments were conducted on TFTR to study the radiation of krypton which will be important for future tokamaks, such as ITER, for the diagnostic of the central ion temperature and for the control of the energy release from the plasma by radiative cooling. The total krypton radiation was monitored, and satellite spectra of Kr XXXV were recorded with a high-resolution crystal spectrometer. Radiative cooling and reduced particle recycling at the plasma edge region were observed, in reasonable agreement with modeling calculations which included radial transport.
Diamagnetic thresholds for sawtooth cycling in tokamak plasmas
Halpern, Federico D.; Luetjens, Hinrich; Luciani, Jean-Francois
2011-10-15
The cycling dynamics of the internal kink mode, which drives sawtooth oscillations in tokamak plasmas, is studied using the three dimensional, non-linear magnetohydrodynamic (MHD) code XTOR-2F [H. Luetjens and J.-F. Luciani, J. Comput. Phys. 229, 8130 (2010)]. It is found that sawtooth cycling, which is characterized by quiescent ramps and fast crashes in the experiment, can be recovered in two-fluid MHD provided that a criterion of diamagnetic stabilization is fulfilled. The simulation results indicate that diamagnetic effects alone may be sufficient to drive sawteeth with complete magnetic reconnection in high temperature Ohmic plasmas.
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.
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.
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
Detached scrape-off layer tokamak plasmas
Kesner, J.
1995-06-01
The equilibrium and stability of scrape-off layer plasmas are considered using a one-dimensional treatment of coupled heat conduction and pressure balance equations. It is found that, for sufficiently low-temperature and high neutral density, a region of greatly reduced power flux to the endplate can be achieved. The plasma in the vicinity of the end wall is characterized by a sharp plasma pressure gradient and a relatively low temperature, 1{lt}{ital T}{sub 0}{lt}10 eV.
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.
Equilibrium calculations for plasma control in CIT (Compact Ignition Tokamak)
Strickler, D.J.; Peng, Y-K.M. . Fusion Engineering Design Center); Pomphrey, N.; Jardin, S.C. . Plasma Physics Lab.)
1990-01-01
The free-boundary equilibrium code VEQ provides equilibrium data that are used by the Tokamak Simulation Code (TSC) in design and analysis of the poloidal field (PF) system for the Compact Ignition Tokamak (CIT). VEQ serves as an important design tool for locating the PF coils and defining coil current trajectories and control systems for TSC. In this report, VEQ and its role in the TSC analysis of the CIT PF system are described. Equilibrium and coil current calculations are discussed, an overview of the CIT PF system is presented, a set of reference equilibria for modeling a complete discharge in CIT is described, and the concept of a plasma shape control matrix is introduced. 9 refs., 8 figs., 7 tabs.
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.
QUIESCENT DOUBLE BARRIER H-MODE PLASMAS IN THE DIII-D TOKAMAK
K.H. BURRELL; M.E. AUSTIN; D.P. BRENNAN; J.C. DeBOO; E.J. DOYLE; C. FENZI; C. FUCHS; P. GOHIL; R.J. GROEBNER; L.L. LAO; T.C. LUCE; M.A. MAKOWSKI; G.R. McKEE; R.A. MOYER; C.C. PETTY; M. PORKOLAB; C.L.RETTIG; T.L. RHODES; J.C. ROST; B.W. STALLARD; E.J. STRAIT; E.J. SYNAKOWSKI; M.R. WADE; J.G. WATKINS; W.P. WEST
2000-11-01
High confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D [J.L. Luxon, et al., Plasma Phys. and Contr. Nucl. Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987) Vol. I, p. 159] this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation which is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 seconds or >25 energy confinement times {tau}{sub E}), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved {beta}{sub N} H{sub 89} = 7 for up to 5 times the {tau}{sub E} of 150 ms. The {beta}{sub N} H{sub 89} values of 7 substantially exceed the value of 4 routinely achieved in standard ELMing H-mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H-mode is possible because of the presence of an edge magnetic hydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.
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
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas.
Green, D L; Berry, L A; Chen, G; Ryan, P M; Canik, J M; Jaeger, E F
2011-09-30
Observations of improved radio frequency (rf) heating efficiency in ITER relevant high-confinement (H-)mode plasmas on the National Spherical Tokamak Experiment are investigated by whole-device linear simulation. The steady-state rf electric field is calculated for various antenna spectra and the results examined for characteristics that correlate with observations of improved or reduced rf heating efficiency. We find that launching toroidal wave numbers that give fast-wave propagation in the scrape-off plasma excites large amplitude (∼kV m(-1)) coaxial standing modes between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggests that these modes are a probable cause of degraded heating efficiency.
The role of plasma rotation on MHD instabilities in tokamaks
Varadarajan, V.; Miley, G.H.
1993-06-01
An improved analysis of the linear stage of the internal kink mode has been developed to include plasma rotation and finite aspect ratio effects. The linear instabiliy growth rates are increased by the plasma rotation. A pseudo-variational, bilinear formalism is used to discretize the linear instability equations; Fourier decomposition is used in the periodic coordinate, and a mixed-finite element procedure is adopted in the radial direction. The numerical studies with the resulting PEST-like code can be used to predict the complex plasma eigenfrequencies. The finite aspect ratio results are similar to the large aspect ratio results for flow instability. The complex instability frequencies found in the ``fishbone`` and TAE modes would be strong determined by the large plasma rotation velocities observed in present-day tokamak devices. These effects could be studied by using the computationally convenient bilinear form derived from the Frieman-Rotenberg equation.
The role of plasma rotation on MHD instabilities in tokamaks
Varadarajan, V.; Miley, G.H.
1993-01-01
An improved analysis of the linear stage of the internal kink mode has been developed to include plasma rotation and finite aspect ratio effects. The linear instabiliy growth rates are increased by the plasma rotation. A pseudo-variational, bilinear formalism is used to discretize the linear instability equations; Fourier decomposition is used in the periodic coordinate, and a mixed-finite element procedure is adopted in the radial direction. The numerical studies with the resulting PEST-like code can be used to predict the complex plasma eigenfrequencies. The finite aspect ratio results are similar to the large aspect ratio results for flow instability. The complex instability frequencies found in the fishbone'' and TAE modes would be strong determined by the large plasma rotation velocities observed in present-day tokamak devices. These effects could be studied by using the computationally convenient bilinear form derived from the Frieman-Rotenberg equation.
Identification of the plasma boundary shape and position in the Damavand tokamak
NASA Astrophysics Data System (ADS)
Rasouli, C.; Abbasi Davani, F.
2017-01-01
A series of experiments and numerical calculations have been done on the Damavand tokamak for accurate determination of equilibrium parameters, such as the plasma boundary position and shape. For this work, the pickup coils of the Damavand tokamak were recalibrated and after that a plasma boundary shape identification code was developed for analyzing the experimental data, such as magnetic probes and coils currents data. The plasma boundary position, shape and other parameters are determined by the plasma shape identification code. A free-boundary equilibrium code was also generated for comparison with the plasma boundary shape identification results and determination of required fields to obtain elongated plasma in the Damavand tokamak.
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.
Simulation of plasma flow in the DIII-D Tokamak
Porter, G. D., LLNL
1998-06-19
The importance of the parallel flow of primary and impurity ions in the Scrape-Off layer (SOL) of divertor tokamaks has been recognized recently. Impurity accumulation on the closed flux surfaces is determined in part by their parallel flow in the SOL. In turn, the parallel transport of the impurity ions is determined in part by drag from the primary ion flow. Measurement of flow in the DIII-D tokamak has begun recently. We describe initial results of modeling plasma ion flow using the 2-D code UEDGE in this paper. We assume the impurity (carbon) arises from chemical and physical sputtering from the walls surrounding the DIII-D plasma. We include six charge states of carbon in our simulations. We make detailed compaison with a multitude of SOL plasma diagnostics, including the flow measurement, to verify the UEDGE physics model. We begin the paper with a brief description of the plasma and neutral models in the UEDGE code in Section 2. We then present initial results of flow simulations and compare them with experimental measurement in Section 3. We conclude with a discussion of the dominant physics processes identified in the modeling in Section 4.
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.
Development of Plasma Equilibrium Response Model for Optimized Plasma Control of KSTAR tokamak
NASA Astrophysics Data System (ADS)
Jeon, Youngmu; Park, Jong-Kyu; Park, Young-Seok; Hwang, Y. S.
2004-11-01
Plasma equilibrium response models for an optimized control system design are developed with KSTAR tokamak configurations. In a simple filament model, plasma column is assumed as a single ring filament with rigid displacements, and constitutes circuits with external conductors (coils, passive plate, and vacuum vessel segments). Perturbed equilibrium response model, based on CREATE-L deformable plasma response model [1], assumes that the plasma evolves through a sequence of MHD equilibria. Prediction characteristics of both models are described in terms of open loop characteristics of vertical motion of plasma, and validated by comparison with TSC (Tokamak Simulation Code) simulations. Additionally, applications of the plasma equilibrium response models to design of optimal plasma controllers are described. [1] R. Albanese, and F. Villone, Nucl. Fusion 38 723 (1998)
Toroidal Alfven Waves in Advanced Tokamaks
NASA Astrophysics Data System (ADS)
Berk, Herbert L.
2003-10-01
In burning plasma experiments, alpha particles have speeds that readily resonate with shear Alfven waves. It is essential to understand this Alfven wave spectrum for toroidal plasma confinement. Most interest has focused on the Toroidal Alfven Eigenmode (TAE), and a method of analysis has been developed to understand the structure of this mode at a flux surface with a given magnetic shear. However, this model fails when the shear is too low or reversed. In this case a new method of analysis is required, which must incorporate novel fluid-like effects from the energetic particles [1] and also include effects that are second order in the inverse toroidal aspect ratio. With this new method [2] we can obtain spectral features that agree with experimental results. In particular, this theory gives an explanation for the so-called Cascade modes that have been observed in JT-60 [3], JET [4], and TFTR [5]. For these Cascade modes, slow upward frequency sweeping is observed, beginning from frequencies below the TAE range but then often blending into the TAE range of frequencies. The theoretical understanding of the Cascades modes has evolved to the point where these modes can be used as a diagnostic "signature" [6] to experimentally optimize the formation of thermal barriers in reversed-shear operation when the minimum q value is an integer. [1] H. L. Berk et al., Phys. Rev. Lett. 87, 185 (2002). [2] B. N. Breizman et al., submitted to Phys. Plasmas (2003). [3] H. Kimura et al., Nucl. Fusion 38, 1303 (1998). [4] S. Sharapov et al., Phys. Lett. A 289, 127 (2001); S. Sharapov, Phys. Plasmas 9, 2027 (2002). [5] R. Nazikian, H. L. Berk, et al., Bull. Am. Phys. Soc. 47, 327 (2002). [6] E. Joffrin et al., Plasma Phys. Contr. Fusion 44, 1739 (2002); E. Joffrin et al., in Proc. 2002 IAEA Fusion Energy Conference, submitted to Nucl. Fusion.
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.
Online Plasma Shape Reconstruction for EAST Tokamak
NASA Astrophysics Data System (ADS)
Luo, Zhengping; Xiao, Bingjia; Zhu, Yingfei; Yang, Fei
2010-08-01
An online plasma shape reconstruction, based on the offline version of the EFIT code and MPI library, can be carried out between two adjacent shots in EAST. It combines online data acquisition, parallel calculation, and data storage together. The program on the master node of the cluster detects the termination of the discharge promptly, reads diagnostic data from the EAST mdsplus server on the completion of data storing, and writes the results onto the EFIT mdsplus server after the calculation is finished. These processes run automatically on a nine-nodes IBM blade center. The total time elapsed is about 1 second to several minutes, depending on the duration of the shot. With the results stored in the mdsplus server, it is convenient for operators and physicists to analyze the behavior of plasma using visualization tools.
Transition to subcritical turbulence in a tokamak plasma
NASA Astrophysics Data System (ADS)
van Wyk, F.; Highcock, E. G.; Schekochihin, A. A.; Roach, C. M.; Field, A. R.; Dorland, W.
2016-12-01
Tokamak turbulence, driven by the ion-temperature gradient and occurring in the presence of flow shear, is investigated by means of local, ion-scale, electrostatic gyrokinetic simulations (with both kinetic ions and electrons) of the conditions in the outer core of the Mega-Ampere Spherical Tokamak (MAST). A parameter scan in the local values of the ion-temperature gradient and flow shear is performed. It is demonstrated that the experimentally observed state is near the stability threshold and that this stability threshold is nonlinear: sheared turbulence is subcritical, i.e. the system is formally stable to small perturbations, but, given a large enough initial perturbation, it transitions to a turbulent state. A scenario for such a transition is proposed and supported by numerical results: close to threshold, the nonlinear saturated state and the associated anomalous heat transport are dominated by long-lived coherent structures, which drift across the domain, have finite amplitudes, but are not volume filling; as the system is taken away from the threshold into the more unstable regime, the number of these structures increases until they overlap and a more conventional chaotic state emerges. Whereas this appears to represent a new scenario for transition to turbulence in tokamak plasmas, it is reminiscent of the behaviour of other subcritically turbulent systems, e.g. pipe flows and Keplerian magnetorotational accretion flows.
Scharer, J.E.
1992-12-31
The research performed under this grant during the past year has been concentrated on the following several key tokamak ICRF (Ion Cyclotron Range of Frequencies) coupling, heating and current drive issues: Efficient coupling during the L- to H- mode transition by analysis and computer simulation of ICRF antennas; analysis of ICRF cavity-backed coil antenna coupling to plasma edge profiles including fast and ion Bernstein wave coupling for heating and current drive; benchmarking the codes to compare with current JET, D-IIID and ASDEX experimental results and predictions for advanced tokamaks such as BPX and SSAT (Steady-State Advanced Tokamak); ICRF full-wave field solutions, power conservation, heating analyses and minority ion current drive; and the effects of fusion alpha particle or ion tail populations on the ICRF absorption. Research progress, publications, and conference and workshop presentations are summarized in this report.
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 anomalous radial transport.
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.
Plasma rotation from momentum transport by neutrals in tokamaks
NASA Astrophysics Data System (ADS)
Omotani, J.; Pusztai, I.; Newton, S.; Fülöp, T.
2016-12-01
Neutral atoms can strongly influence the intrinsic rotation and radial electric field at the tokamak edge. Here, we present a framework to investigate these effects when the neutrals dominate the momentum transport. We explore the parameter space numerically, using highly flexible model geometries and a state of the art kinetic solver. We find that the most important parameters controlling the toroidal rotation and electric field are the major radius where the neutrals are localized and the plasma collisionality. This offers a means to influence the rotation and electric field by, for example, varying the radial position of the X-point to change the major radius of the neutral peak.
Nearly axisymmetric hot plasmas in a highly rippled tokamak
NASA Astrophysics Data System (ADS)
Bellan, Paul
2002-11-01
Tokamak ohmic heating current flowing along toroidally rippled flux surfaces results in a poloidal torque. Since pressure gradients cannot offset torques, the torque drives plasma flows which convect plasma toroidally from ripple necks (high B_pol^2) to ripple bulges (low B_pol^2). Stagnation of the oppositely directed toroidal flows at the ripple bulges thermalizes the directed flow velocity ˜ B_pol/μ_0ρ , giving β _pol ˜1. These flows also convect frozen-in poloidal field lines which accumulate at the bulges enhancing the pinch force there and so reducing the bulge. Thus, a nearly axisymmetric β_pol ˜1 equilibrium is achieved using only a few TF coils. Particles bouncing in step between approaching flows will be Fermi accelerated to form a high energy tail. The ST tokamak magnetic mountain experiment [1] showed that, compared to a 1.8% ripple configuration, a 28% ripple configuration had four times the neutron production, and only a modest degradation of overall confinement; the former is consistent with the notion of Fermi acceleration of particles bouncing between colliding toroidal flows and the latter is consistent with ripple reduction due to toroidal convection of poloidal field lines. [1] W. Stodiek et al, Proc. 4th Intl. Conf. Plasma Phys. and Contr. Nuc. Fusion Res., (Madison, 1971), Vol. 1, p. 465
Turbulent-driven intrinsic rotation in tokamak plasmas
NASA Astrophysics Data System (ADS)
Barnes, Michael; Parra, Felix; Lee, Jungpyo; Belli, Emily; Nave, Filomena; White, Anne
2013-10-01
Tokamak plasmas are routinely observed to rotate even in the absence of an externally applied torque. This ``intrinsic'' rotation exhibits several robust features, including rotation reversals with varying plasma density and current and rotation peaking at the transition from low confinement to high confinement regimes. Conservation of toroidal angular momentum dictates that the intrinsic rotation is determined by momentum redistribution within the plasma, which is dominated by turbulent transport. The turbulent momentum transport, and thus the intrinsic rotation profile, is driven by formally small effects that are usually neglected. We present a gyrokinetic theory that makes use of the smallness of the poloidal to total magnetic field ratio to self-consistently include the dominant effects driving intrinsic turbulent momentum transport in tokamaks. These effects (including slow radial profile variation, slow poloidal turbulence variation, and diamagnetic corrections to the equilibrium Maxwellian) have now been implemented in the local, delta-f gyrokinetic code GS2. We describe important features of the numerical implementation and show numerical results on intrinsic momentum transport that are qualitatively consistent with experimental rotation reversals.
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.
On plasma rotation induced by waves in tokamaks
Guan, Xiaoyin; Dodin, I. Y.; Fisch, N. J.; Qin, Hong; 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.
A divertor plasma configuration design method for tokamaks
NASA Astrophysics Data System (ADS)
Guo, Yong; Xiao, Bing-Jia; Liu, Lei; Yang, Fei; Wang, Yuehang; Qiu, Qinglai
2016-11-01
The efficient and safe operation of large fusion devices strongly relies on the plasma configuration inside the vacuum chamber. It is important to construct the proper plasma equilibrium with a desired plasma configuration. In order to construct the target configuration, a shape constraint module has been developed in the tokamak simulation code (TSC), which controls the poloidal flux and the magnetic field at several defined control points. It is used to construct the double null, lower single null, and quasi-snowflake configurations for the required target shape and calculate the required PF coils current. The flexibility and practicability of this method have been verified by the simulated results. Project supported by the National Magnetic Confinement Fusion Research Program of China (Grant Nos. 2014GB103000 and 2014GB110003), the National Natural Science Foundation of China (Grant Nos. 11305216, 11305209, and 11375191), and External Cooperation Program of BIC, Chinese Academy of Sciences (Grant No. GJHZ201303).
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
Magnetic confinement experiment -- 1: Tokamaks
Goldston, R.J.
1994-12-31
This report reviews presentations made at the 15th IAEA Conference on Plasma Physics and Controlled Nuclear Fusion on experimental tokamak physics, particularly on advances in core plasma physics, divertor and edge physics, heating and current drive, and tokamak concept optimization.
Numerical investigation of plasma-wall interaction during burst of ELMs in a tokamak device
NASA Astrophysics Data System (ADS)
Ou, Jing; Lin, Binbin; Zhao, Xiaoyun
2017-01-01
In the tokamak high confinement mode (H-mode), the transient heat fluxes caused by edge localized modes (ELMs) will eventually irradiate the plasma-facing components and may erode, even melt them. To study the performance of the plasma-wall interaction during ELMs, interactions among heat flux from plasma, surface temperature, and electron emission are simulated by considering self-consistency among plasma transport in the sheath, deposited heat at the wall, and material thermal response for carbon (C) and tungsten (W) walls. It is found that the sheath structure determines the surface temperature, which may in turn influence on the sheath. A large amount of electron emission can change the heat load from the plasma to the material surface due to the variation of the ELMs-induced electron temperature and the surface temperature. During the burst of ELMs, the surface temperature rises rapidly at first and then reaches a saturation state with a certain range of fluctuation. The development of these processes depends strongly on the characteristic of ELMs, deposited heat at the wall, and material properties. Simulation results also show that the erosion of the Experimental Advanced Superconducting Tokamak (EAST) divertor target is of no concern in H-mode operation with ELMs for the current and possible future operation parameters.
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.
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.
Quiescent Double Barrier H-Mode Plasmas in the DIII-D Tokamak
Burrell, K H; Austin, M E; Brennan, D P; DeBoo, J C; Doyle, E J; Fenzi, C; Fuchs, C; Gohil, P; Greenfield, C M; Groebner, R J; Lao, L L; Luce, T C; Makowski, M A; McKee, G R; Moyer, R A; Petty, C C; Porkolab, M; Rettig, C L; Rhodes, T L; Rost, J C; Stallard, B W; Strait, E J; Synakowski, E J; Wade, M R; Watkins, J G; West, W P
2000-11-01
High confinement (H-mode) operation is the choice for next-step tokamak devices based either on conventional or advanced tokamak physics. This choice, however, comes at a significant cost for both the conventional and advanced tokamaks because of the effects of edge localized modes (ELMs). ELMs can produce significant erosion in the divertor and can affect the beta limit and reduced core transport regions needed for advanced tokamak operation. Experimental results from DIII-D this year have demonstrated a new operating regime, the quiescent H-mode regime, which solves these problems. We have achieved quiescent H-mode operation which is ELM-free and yet has good density and impurity control. In addition, we have demonstrated that an internal transport barrier can be produced and maintained inside the H-mode edge barrier for long periods of time (>3.5 seconds or >25 energy confinement times {tau}{sub E}), yielding a quiescent double barrier regime. By slowly ramping the input power, we have achieved {beta}{sub N} H89 = 7 for up to 5 times the {tau}{sub E} of 150 ms. The {beta}{sub N} H89 values of 7 substantially exceed the value of 4 routinely achieved in standard ELMing H-mode. The key factors in creating the quiescent H-mode operation are neutral beam injection in the direction opposite to the plasma current (counter injection) plus cryopumping to reduce the density. Density and impurity control in the quiescent H-mode is possible because of the presence of an edge magnetic hydrodynamic (MHD) oscillation, the edge harmonic oscillation, which enhances the edge particle transport while leaving the energy transport unaffected.
Predicting high harmonic ion cyclotron heating efficiency in Tokamak plasmas
Green, David L; Jaeger, E. F.; Berry, Lee A; Chen, Guangye; Ryan, Philip Michael; Canik, John
2011-01-01
Observations of improved radio frequency (RF) heating efficiency in high-confinement (H-) mode plasmas on the National Spherical Tokamak Experiment (NSTX) are investigated by whole-device linear simulation. We present the first full-wave simulation to couple kinetic physics of the well confined core plasma to the poorly confined scrape-off plasma. The new simulation is used to scan the launched fast-wave spectrum and examine the steady-state electric wave field structure for experimental scenarios corresponding to both reduced, and improved RF heating efficiency. We find that launching toroidal wave-numbers that required for fast-wave propagation excites large amplitude (kVm 1 ) coaxial standing modes in the wave electric field between the confined plasma density pedestal and conducting vessel wall. Qualitative comparison with measurements of the stored plasma energy suggest these modes are a probable cause of degraded heating efficiency. Also, the H-mode density pedestal and fast-wave cutoff within the confined plasma allow for the excitation of whispering gallery type eigenmodes localised to the plasma edge.
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.
Halo current diagnostic system of experimental advanced superconducting tokamak
NASA Astrophysics Data System (ADS)
Chen, D. L.; Shen, B.; Granetz, R. S.; Sun, Y.; Qian, J. P.; Wang, Y.; Xiao, B. J.
2015-10-01
The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.
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.
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.
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.
Impurity poloidal asymmetries and plasma rotation in the PDX Tokamak
NASA Astrophysics Data System (ADS)
Brau, K.
Vertical poloidal asymmetries of carbon and oxygen in the PDX Tokamak were monitored under a variety of discharge conditions in circular plasmas. Near the edge of the plasma and in the region beyond the limiter, the asymmetries appear to be caused by local impurity recycling, variations in the length of the emitting region, and effects due to vertical ion drifts. In the case of C V impurities, the sign and magnitude of the asymmetry is in qualitative agreement with the predictions of a quasi-neoclassical fluid model of impurity transport. A two dimensional computer code is used to simulate different models of poloidal asymmetries, including: (1) poloidally asymmetric source function, (2) charge exchange recombination with neutral hydrogen, (3) poloidally asymmetric electron ensity and temperature profiles, (4) poloidally varying anomalous radial diffusion coefficient, and (5) the quasi-neoclassical fluid model.
Remote network control plasma diagnostic system for Tokamak T-10
NASA Astrophysics Data System (ADS)
Troynov, V. I.; Zimin, A. M.; Krupin, V. A.; Notkin, G. E.; Nurgaliev, M. R.
2016-09-01
The parameters of molecular plasma in closed magnetic trap is studied in this paper. Using the system of molecular diagnostics, which was designed by the authors on the «Tokamak T-10» facility, the radiation of hydrogen isotopes at the plasma edge is investigated. The scheme of optical radiation registration within visible spectrum is described. For visualization, identification and processing of registered molecular spectra a new software is developed using MatLab environment. The software also includes electronic atlas of electronic-vibrational-rotational transitions for molecules of protium and deuterium. To register radiation from limiter cross-section a network control system is designed using the means of the Internet/Intranet. Remote control system diagram and methods are given. The examples of web-interfaces for working out equipment control scenarios and viewing of results are provided. After test run in Intranet, the remote diagnostic system will be accessible through Internet.
NASA Astrophysics Data System (ADS)
Shi, Yuejiang; Fu, Jia; Li, Jiahong; Yang, Yu; Wang, Fudi; Li, Yingying; Zhang, Wei; Wan, Baonian; Chen, Zhongyong
2010-03-01
The synchrotron radiation originated from the energetic runaway electrons has been measured by a visible complementary metal oxide semiconductor camera working in the wavelength ranges of 380-750 nm in the Experimental Advanced Superconducting Tokamak [H. Q. Liu et al., Plasma Phys. Contr. Fusion 49, 995 (2007)]. With a tangential viewing into the plasma in the direction of electron approach on the equatorial plane, the synchrotron radiation from the energetic runaway electrons was measured in full poloidal cross section. The synchrotron radiation diagnostics provides a direct pattern of the runaway beam inside the plasma. The energy and pitch angle of runaway electrons have been obtained according to the synchrotron radiation pattern. A stable shell shape of synchrotron radiation has been observed in a few runaway discharges.
Shi Yuejiang; Fu Jia; Li Jiahong; Yang Yu; Wang Fudi; Li Yingying; Zhang Wei; Wan Baonian; Chen Zhongyong
2010-03-15
The synchrotron radiation originated from the energetic runaway electrons has been measured by a visible complementary metal oxide semiconductor camera working in the wavelength ranges of 380-750 nm in the Experimental Advanced Superconducting Tokamak [H. Q. Liu et al., Plasma Phys. Contr. Fusion 49, 995 (2007)]. With a tangential viewing into the plasma in the direction of electron approach on the equatorial plane, the synchrotron radiation from the energetic runaway electrons was measured in full poloidal cross section. The synchrotron radiation diagnostics provides a direct pattern of the runaway beam inside the plasma. The energy and pitch angle of runaway electrons have been obtained according to the synchrotron radiation pattern. A stable shell shape of synchrotron radiation has been observed in a few runaway discharges.
Shi, Yuejiang; Fu, Jia; Li, Jiahong; Yang, Yu; Wang, Fudi; Li, Yingying; Zhang, Wei; Wan, Baonian; Chen, Zhongyong
2010-03-01
The synchrotron radiation originated from the energetic runaway electrons has been measured by a visible complementary metal oxide semiconductor camera working in the wavelength ranges of 380-750 nm in the Experimental Advanced Superconducting Tokamak [H. Q. Liu et al., Plasma Phys. Contr. Fusion 49, 995 (2007)]. With a tangential viewing into the plasma in the direction of electron approach on the equatorial plane, the synchrotron radiation from the energetic runaway electrons was measured in full poloidal cross section. The synchrotron radiation diagnostics provides a direct pattern of the runaway beam inside the plasma. The energy and pitch angle of runaway electrons have been obtained according to the synchrotron radiation pattern. A stable shell shape of synchrotron radiation has been observed in a few runaway discharges.
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)
A control-oriented model of the current profile in tokamak plasma
NASA Astrophysics Data System (ADS)
Witrant, E.; Joffrin, E.; Brémond, S.; Giruzzi, G.; Mazon, D.; Barana, O.; Moreau, P.
2007-07-01
This paper proposes a control-oriented approach to the tokamak plasma current profile dynamics. It is established based on a consistent set of simplified relationships, in particular for the microwave current drive sources, rather than exact physical modelling. Assuming that a proper model for advanced control schemes can be established using the so-called cylindrical approximation and neglecting the diamagnetic effects, we propose a model that focuses on the flux diffusion (from which the current profile is inferred). Its inputs are some real-time measurements available on modern tokamaks and the effects of some major actuators, such as the magnetic coils, lower hybrid (LHCD), electron and ion cyclotron frequency (ECCD and ICRH) systems, are particularly taken into account. More precisely, the non-inductive current profile sources are modelled as 3-parameters functions of the control inputs derived either from approximate theoretical formulae for the ECCD and bootstrap terms or from experimental scaling laws specifically developed from hard x-ray Tore Supra data for the LHCD influence. The use of scaling laws in this model reflects the fact that the operation of future reactors will certainly depend upon a great number of scaling laws and specific engineering parameters. The discretization issues are also specifically addressed, to ensure robustness with respect to discretisation errors and the efficiency (in terms of computation time) of the associated algorithm. This model is compared with experimental results and the CRONOS solver for tore supra tokamak.
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.
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
Applying the new HIT results to tokamak and solar plasmas
NASA Astrophysics Data System (ADS)
Jarboe, Thomas; Sutherland, Derek; Hossack, Aaron; Nelson, Brian; Morgan, Kyle; Chris, Hansen; Benedett, Thomas; Everson, Chris; Penna, James
2016-10-01
Understanding sustainment of stable equilibria with helicity injection in HIT-SI has led to a simple picture of several tokamak features. Perturbations cause a viscous-like force on the current that flattens the λ profile, which sustains and stabilizes the equilibrium. An explanation of the mechanism is based on two properties of stable, ideal, two-fluid, magnetized plasma. First, the electron fluid is frozen to magnetic fields and, therefore, current flow is also magnetic field flow. Second, for a stable equilibrium the structure perpendicular to the flux surface resists deformation. Thus toroidal current is from electrons frozen in nested, rotating resilient flux surfaces. Only symmetric flux surfaces allow free differential current flow. Perturbations cause interference of the flux surfaces. Thus, perturbations cause forces that oppose differential electron rotation and forced differential flow produces a symmetrizing force against perturbations and instability. This mechanism can explain the level of field error that spoils tokamak performance and the rate of poloidal flux loss in argon-induced disruptions in DIII-D. This new understanding has led to an explanation of the source of the solar magnetic fields and the power source for the chromosphere, solar wind and corona. Please place in spheromak and FRC section with other HIT posters.
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.
Real-time extraction of plasma equilibrium parameters in KSTAR tokamak using statistical methods
NASA Astrophysics Data System (ADS)
Na, Yong-Su; Jeon, Young-Mu; Hong, S. H.; Hwang, Y. S.
2001-02-01
To improve inherent shortcomings of statistical methods and apply them to the extraction of plasma equilibrium parameters in a fast timescale for real-time plasma control, new concepts of statistical methods such as principal component analysis-based neural network (NN), functional parametrization (FP)-based NN and double network are introduced by modifying NN and FP. These new methods are benchmarked and compared with the conventional techniques of NN and FP in a simple single-filament system. As a result of their applications to identification of plasma equilibrium parameters in the Korea Superconducting Tokamak Advanced Research tokamak, particularly, the double network concept among them has successfully achieved the improvement of drawbacks in the conventional methods. It is shown that more reliable results from the double network method can be obtained by combining several different statistical treatments as a primary network. Even in the case of nonoptimized methods united as a primary network, quite acceptable results can be achieved in the double network method.
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.
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.
Spectra of heliumlike krypton from tokamak fusion test reactor plasmas
Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M.; Beiersdorfer, P.; Osterheld, A.; Smith, A.; Fraenkel, B.
1993-04-01
Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 {Angstrom} including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport.
Spectra of heliumlike krypton from tokamak fusion test reactor plasmas
Bitter, M.; Hsuan, H.; Bush, C.; Cohen, S.; Cummings, C.J.; Grek, B.; Hill, K.W.; Schivell, J.; Zarnstorff, M. . Plasma Physics Lab.); Beiersdorfer, P.; Osterheld, A. ); Smith, A. ); Fraenkel, B. )
1993-04-01
Krypton has been injected into ohmically-heated TFTR plasmas with peak electron temperatures of 6 key to study the effects of krypton on the plasma performance and to investigate the emitted krypton line radiation, which is of interest for future-generation tokamaks such as ITER, both as a diagnostic of the central ion temperature and for the control of energy release from the plasma by radiative cooling. The emitted radiation was monitored with a bolometer array, an X-ray pulse height analysis system, and a high-resolution Johann-type crystal spectrometer; and it was found to depend very sensitively on the electron temperature profile. Satellite spectra of heliumlike krypton, KrXXXV, near 0.95 [Angstrom] including lithiumlike, berylliumlike and boronlike features were recorded in second order Bragg reflection. Radiative cooling and reduced particle recycling at the plasma edge region were observed as a result of the krypton injection for all investigated discharges. The observations are in reasonable agreement with modeling calculations of the krypton ion charge state distribution including radial transport.
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.; ...
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
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)].
Continuum kinetic modeling of the tokamak plasma edge
Dorf, M. A.; Dorr, M.; Rognlien, T.; ...
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
Turbulent transport of alpha particles in tokamak plasmas
NASA Astrophysics Data System (ADS)
Croitoru, A.; Palade, D. I.; Vlad, M.; Spineanu, F.
2017-03-01
We investigate the \\boldsymbol{E}× \\boldsymbol{B} diffusion of fusion born α particles in tokamak plasmas. We determine the transport regimes for a realistic model that has the characteristics of the ion temperature gradient (ITG) or of the trapped electron mode (TEM) driven turbulence. It includes a spectrum of potential fluctuations that is modeled using the results of the numerical simulations, the drift of the potential with the effective diamagnetic velocity and the parallel motion. Our semi-analytical statistical approach is based on the decorrelation trajectory method (DTM), which is adapted to the gyrokinetic approximation. We obtain the transport coefficients as a function of the parameters of the turbulence and of the energy of the α particles. According to our results, significant turbulent transport of the α particles can appear only at energies of the order of 100 KeV. We determine the corresponding conditions.
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.
Study of Plasma MHD Equilibrium in HL-2A Tokamak
NASA Astrophysics Data System (ADS)
He, Hongda; Zhang, Jinhua; Dong, Jiaqi; Li, Qiang
2006-07-01
EFIT(Equilibrium Fitting) code is successfully transplanted for HL-2A tokamak parameters. The evolution of the equilibrium configurations for shot 2898 is simulated. It is shown that the discharge starts with a limiter configuration and develops into a divertor configuration gradually after t = 200 ms. The latter lasts until the end of the discharge at t = 900 ms. The evolution of the major plasma shape parameters such as the boundary magnetic fluxes, the positions of the x-point and magnetic axis, and the minor radii obtained are calculated and compared with the experimental results. The agreement between the simulation and experiments are shown to be reasonable. The possibility for discharge quality improvement is discussed.
Plasma position control in the STOR-M tokamak: A fuzzy logic approach
NASA Astrophysics Data System (ADS)
Morelli, Jordan Edwin
Adequate control of the position of the plasma column within the STOR-M tokamak is a chief requirement in order for experimental quality discharges to be obtained. Optimal control over tokamak discharge parameters, including the plasma position, is very difficult to achieve. This is due in large part to the difficulty in modelling the tokamak discharge parameters, as they are highly nonlinear and time varying in nature. The difficulty of modelling the tokamak discharge parameters suggests that a control system, such as a fuzzy logic based controller, which does not require a system model may be well suited to the control of fusion plasma. In order to improve the quality of control over the plasma position within the STOR-M tokamak, the existing analog PID controller was modified. These modifications facilitate the application of a digital controller by a personal computer via the Advantech PCL-711B data acquisition card. The performance of the modified plasma position controller and an Arbitrary Signal Generator developed by the author was evaluated. This modified plasma position controller was applied successfully to the STOR-M tokamak during both normal mode and A.C. mode operation. In both cases, the modified controller provided adequate control over the position of the plasma column within the discharge chamber. Furthermore, the modified controller was more convenient to optimize than the original, existing analog PID controller. By taking advantage of the modifications that were made to the plasma position controller, a fuzzy logic controller was developed by the author. The fuzzy logic based plasma position controller was also successfully applied to the STOR-M tokamak during both normal mode and A.C. operation. The fuzzy controller was demonstrated to reliably provide a higher degree of control over the position of the plasma column within the STOR-M tokamak than the modified PID controller.
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
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.
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.
Design of vibration compensation interferometer for Experimental Advanced Superconducting Tokamak
NASA Astrophysics Data System (ADS)
Yang, Y.; Li, G. S.; Liu, H. Q.; Jie, Y. X.; Ding, W. X.; Brower, D. L.; Zhu, X.; Wang, Z. X.; Zeng, L.; Zou, Z. Y.; Wei, X. C.; Lan, T.
2014-11-01
A vibration compensation interferometer (wavelength at 0.532 μm) has been designed and tested for Experimental Advanced Superconducting Tokamak (EAST). It is designed as a sub-system for EAST far-infrared (wavelength at 432.5 μm) poloarimeter/interferometer system. Two Acoustic Optical Modulators have been applied to produce the 1 MHz intermediate frequency. The path length drift of the system is lower than 2 wavelengths within 10 min test, showing the system stability. The system sensitivity has been tested by applying a periodic vibration source on one mirror in the system. The vibration is measured and the result matches the source period. The system is expected to be installed on EAST by the end of 2014.
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.
Optical boundary reconstruction of tokamak plasmas for feedback control of plasma position and shape
NASA Astrophysics Data System (ADS)
Hommen, G.; de Baar, M.; Nuij, P.; McArdle, G.; Akers, R.; Steinbuch, M.
2010-11-01
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.
Hommen, G; de Baar, M; Nuij, P; McArdle, G; Akers, R; Steinbuch, M
2010-11-01
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.
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.
Investigations in the Nonlinear Dynamics of Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Lebedev, Vladimir Borisov
1995-01-01
Analytical and numerical investigations of modulational interaction between drift waves and trapped ion convective cells as well as a simple model of Edge Localized Mode (ELM) phenomena in tokamak plasma are presented in this dissertation. There are two main parts. In the first part, the linear and nonlinear dynamics of modulational interaction between small scale drift waves and large scale trapped ion convective cells are investigated. A set of envelope equations describing this interaction has been derived and analyzed, both numerically and analytically. The growth rate of modulational instability is determined by spectral properties of drift waves and can exceed the linear growth rate of the trapped ion mode. An anisotropic spectrum of drift waves is always modulationally unstable. For very short wavelength drift waves with k| rho_{s} >= 1, the interaction results in a universal final state of thin anisotropic dipole convective cells which trap the drift waves. The spatial orientation of the convective cell pattern is determined by drift wave spectrum anisotropy and propagation direction. In the presence of a sheared magnetic field the modulational growth rate becomes intrinsically anisotropic, on account of the modified radial structure of drift waves. In the second part, a simple, low-dimensional model of Edge Localized Mode phenomena is presented. ELM dynamics are determined by the interaction of few basic processes at the edge of tokamak plasma, these include: the evolution of magnetohydrodynamic (MHD) pressure gradient driven instabilities, the L-H transition, the fueling of the edge by neutral particles, and edge heating by thermal flux from the core plasma. In the parameter regime characteristic of an H-mode plasma, the model exhibits a transition to stationary relaxation oscillations (i.e. stable limit cycle behavior) corresponding to ELMs. The dependence of ELM frequency, amplitude etc. on the heating power P_{in} and other control parameters is
Emission in the 50-80 A region from highly ionized silver in PLT tokamak plasmas
Schwob, J.L.; Wouters, A.; Suckewer, S.; Cohen, S.A.; Finkenthal, M.
1985-09-01
The spectrum of silver emitted by Princeton Large Torus (PLT) tokamak plasmas has been recorded in the 25 to 150 A region by a multichannel time-resolving grazing-incidence spectrometer. Silver atoms have been introduced in the tokamak plasma using the laser blow-off technique. For the first time, lines emitted within the 3p-3d transitions of Ag XXIX, Ag XXX, and Ag XXXI ions, between 50 and 80 A, have been identified.
Runaway electrons and mitigation studies in MST tokamak plasmas
NASA Astrophysics Data System (ADS)
Goetz, J. A.; Chapman, B. E.; Almagri, A. F.; Cornille, B. S.; Dubois, A.; McCollam, K. J.; Munaretto, S.; Sovinec, C. R.
2016-10-01
Studies of runaway electrons generated in low-density MST tokamak plasmas are being undertaken. The plasmas have Bt <= 0.14 T, Ip <= 50 kA, q (a) = 2.2 , and an electron density and temperature of about 5 ×1017m-3 and 150 eV. Runaway electrons are detected via x-ray bremsstrahlung emission. The density and electric field thresholds for production and suppression have been previously explored with variations in gas puffing for density control. Runaway electrons are now being probed with resonant magnetic perturbations (RMP's). An m = 3 RMP strongly suppresses the runaway electrons and initial NIMROD modeling shows that this may be due to degradation of flux surfaces. The RMP is produced by a poloidal array of 32 saddle coils at the narrow vertical insulated cut in MST's thick conducting shell, with each RMP having a single m but a broad n spectrum. While a sufficiently strong m = 3 RMP suppresses the runaway electrons, an RMP with m = 1 and comparable amplitude has little effect. The impact of the RMP's on the magnetic topology of these plasmas is being studied with the nonlinear MHD code NIMROD. With an m = 3 RMP, stochasticity is introduced in the outer third of the plasma but no such flux surface degradation is observed with an m = 1 RMP. NIMROD also predicts regularly occurring MHD activity similar to that observed in the experiment. These studies have also been done in q (a) = 2.7 plasmas and analysis and modeling is ongoing. This work supported by USDoE.
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
Propagation of global shear Alfven waves in gyrokinetic tokamak plasmas
NASA Astrophysics Data System (ADS)
Nishimura, Y.; Lin, Z.; Holod, I.; Chen, L.; Decyk, V.; Klasky, S.; Ma, K.; Adams, M.; Ethier, S.; Hahm, T.; Lee, W.; Lewandowski, J.; Rewoldt, G.; Wang, W.
2006-04-01
Employing the electromagnetic gyrokinetic simulation models, Alfven wave dynamics in global tokamak geometry is studied. Based on a small parameter expansion by the square-root of the electron-ion mass ratio, the fluid-kinetic hybrid electron model solves the adiabatic response in the lowest order and solves the kinetic response in the higher orders. We verify the propagation of shear Alfven waves in the absence of drives or damping mechanisms by perturbing the magnetic field lines at t=0 in a global eigenmode structure. The Alfven wave experiences continuum damping. In the presence of energetic particles, excitations of toroidal Alfven eigenmode (TAE) is expected within the frequency gap. With the ηi gradient drive, at a critical β value, the kinetic ballooning mode (KBM) is excited below the ideal MHD limit. W.W.Lee et al., Phys. Plasmas 8, 4435 (2001). Z.Lin and L.Chen, Phys. Plasmas 8, 1447 (2001). J.A.Tataronis and W. Grossman, Z. Phys. 14, 203 (1973). C.Z.Cheng, L.Chen, and M.S.Chance, Ann.Phys. 161, 21 (1984). C.Z.Cheng, Nucl. Fusion 22, 773 (1982).
Perpendicular dynamics of runaway electrons in tokamak plasmas
Fernandez-Gomez, I.; Martin-Solis, J. R.; Sanchez, R.
2012-10-15
In this paper, it will be shown that the runaway phenomenon in tokamak plasmas cannot be reduced to a one-dimensional problem, based on the competence between electric field acceleration and collisional friction losses in the parallel direction. A Langevin approach, including collisional diffusion in velocity space, will be used to analyze the two-dimensional runaway electron dynamics. An investigation of the runaway probability in velocity space will yield a criterion for runaway, which will be shown to be consistent with the results provided by the more simple test particle description of the runaway dynamics [Fuchs et al., Phys. Fluids 29, 2931 (1986)]. Electron perpendicular collisional scattering will be found to play an important role, relaxing the conditions for runaway. Moreover, electron pitch angle scattering perpendicularly broadens the runaway distribution function, increasing the electron population in the runaway plateau region in comparison with what it should be expected from electron acceleration in the parallel direction only. The perpendicular broadening of the runaway distribution function, its dependence on the plasma parameters, and the resulting enhancement of the runaway production rate will be discussed.
Nonlinear transport processes in tokamak plasmas. I. The collisional regimes
NASA Astrophysics Data System (ADS)
Sonnino, Giorgio; Peeters, Philippe
2008-06-01
An application of the thermodynamic field theory (TFT) to transport processes in L-mode tokamak plasmas is presented. The nonlinear corrections to the linear ("Onsager") transport coefficients in the collisional regimes are derived. A quite encouraging result is the appearance of an asymmetry between the Pfirsch-Schlüter (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 102. 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 102 when the nonlinear contributions are duly taken into account but, there is still a factor of 102 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.
Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)
Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.
1994-07-01
Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES&H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power.
NASA Astrophysics Data System (ADS)
Nam, Y. U.; Cheon, M. S.; Kwon, M.; Hwang, Y. S.
2003-03-01
A simple single-channel horizontal millimeter-wave interferometer has been designed for plasma electron density measurements on the Korea Superconducting Tokamak Advanced Research (KSTAR). To measure line integrated plasma densities of 2×1019 m-2 in the initial phase of the KSTAR, Gunn oscillator frequency of 280 GHz has been chosen to optimize errors due to both vibration on the beam path and refraction in the plasma. To reduce the free propagation length of the probing beam and to obtain small beam width on the vacuum windows, a retractable cassette system for deep positioning of the diagnostic system has been designed, where microwave parts are located as close as possible to the tokamak with a shielding box. A beam focusing system with concave reflecting mirrors has been designed on the cassette and on the inner wall of the tokamak to reduce beam losses and to minimize beam width in the plasma. The estimated total transmission loss is about 25 dB, and beam widths are reduced significantly in the range of 20-50 mm.
Time parallelization of advanced operation scenario simulations of ITER plasma
Samaddar, D.; Casper, T. A.; Kim, S. H.; Berry, Lee A; Elwasif, Wael R; Batchelor, Donald B; Houlberg, Wayne A
2013-01-01
This work demonstrates that simulations of advanced burning plasma operation scenarios can be successfully parallelized in time using the parareal algorithm. CORSICA - an advanced operation scenario code for tokamak plasmas is used as a test case. This is a unique application since the parareal algorithm has so far been applied to relatively much simpler systems except for the case of turbulence. In the present application, a computational gain of an order of magnitude has been achieved which is extremely promising. A successful implementation of the Parareal algorithm to codes like CORSICA ushers in the possibility of time efficient simulations of ITER plasmas.
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)/
Theory-based scaling of the SOL width in circular limited tokamak plasmas
NASA Astrophysics Data System (ADS)
Halpern, F. D.; Ricci, P.; Labit, B.; Furno, I.; Jolliet, S.; Loizu, J.; Mosetto, A.; Arnoux, G.; Gunn, J. P.; Horacek, J.; Kočan, M.; LaBombard, B.; Silva, C.; Contributors, JET-EFDA
2013-12-01
A theory-based scaling for the characteristic length of a circular, limited tokamak scrape-off layer (SOL) is obtained by considering the balance between parallel losses and non-linearly saturated resistive ballooning mode turbulence driving anomalous perpendicular transport. The SOL size increases with plasma size, resistivity, and safety factor q. The scaling is verified against flux-driven non-linear turbulence simulations, which reveal good agreement within a wide range of dimensionless parameters, including parameters closely matching the TCV tokamak. An initial comparison of the theory against experimental data from several tokamaks also yields good agreement.
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.
Analog integrator for the Korea superconducting tokamak advanced research magnetic diagnostics
Bak, J. G.; Lee, S. G.; Son, D.; Ga, E. M.
2007-04-15
An analog integrator, which automatically compensates an integrating drift, has been developed for the magnetic diagnostics in the Korea superconducting tokamak advanced research (KSTAR). The compensation of the drift is done by the analog to digital converter-register-digital to analog converter in the integrator. The integrator will be used in the equilibrium magnetic field measurements by using inductive magnetic sensors during a plasma discharge in the KSTAR machine. Two differential amplifiers are added to the signal path between each magnetic sensor and the integrator in order to improve the performance of the integrator because a long signal cable of 100 m will be used for the measurement in the KSTAR machine. In this work, the characteristics of the integrator with two differential amplifiers are experimentally investigated.
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.
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.
Analog integrator for the Korea superconducting tokamak advanced research magnetic diagnostics.
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.
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.
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.
NASA Astrophysics Data System (ADS)
Lampert, M.; Anda, G.; Czopf, A.; Erdei, G.; Guszejnov, D.; Kovácsik, Á.; Pokol, G. I.; Réfy, D.; Nam, Y. U.; Zoletnik, S.
2015-07-01
A novel beam emission spectroscopy observation system was designed, built, and installed onto the Korea Superconducting Tokamak Advanced Research tokamak. The system is designed in a way to be capable of measuring beam emission either from a heating deuterium or from a diagnostic lithium beam. The two beams have somewhat complementary capabilities: edge density profile and turbulence measurement with the lithium beam and two dimensional turbulence measurement with the heating beam. Two detectors can be used in parallel: a CMOS camera provides overview of the scene and lithium beam light intensity distribution at maximum few hundred Hz frame rate, while a 4 × 16 pixel avalanche photo-diode (APD) camera gives 500 kHz bandwidth data from a 4 cm × 16 cm region. The optics use direct imaging through lenses and mirrors from the observation window to the detectors, thus avoid the use of costly and inflexible fiber guides. Remotely controlled mechanisms allow adjustment of the APD camera's measurement location on a shot-to-shot basis, while temperature stabilized filter holders provide selection of either the Doppler shifted deuterium alpha or lithium resonance line. The capabilities of the system are illustrated by measurements of basic plasma turbulence properties.
Linear gyrokinetic theory for kinetic magnetohydrodynamic eigenmodes in tokamak plasmas
NASA Astrophysics Data System (ADS)
Qin, H.; Tang, W. M.; Rewoldt, G.
1999-06-01
A two-dimensional (2D) numerical solution method is developed for the recently derived linear gyrokinetic system which describes arbitrary wavelength electromagnetic perturbations in tokamak plasmas. The system consists of the gyrokinetic equation, the gyrokinetic Poisson equation, and the gyrokinetic moment equation. Since familiar magnetohydrodynamic (MHD) results can be recovered entirely from this gyrokinetic model, and all interesting kinetic effects are intrinsically included, this gyrokinetic system offers an approach for kinetic MHD phenomena which is more rigorous, self-consistent, and comprehensive than the previous hybrid models. Meanwhile, drift type microinstabilities can be also investigated systematically in this theoretical framework. The linear gyrokinetic equation is solved for the distribution function in terms of the perturbed fields by integrating along unperturbed particle orbits. The solution is substituted back into the gyrokinetic moment equation and the gyrokinetic Poisson equation. When the boundary conditions are incorporated, an eigenvalue problem is formed. The resulting numerical code, KIN-2DEM, is applied to kinetic ballooning modes, internal kink modes, and toroidal Alfvén eigenmodes (TAEs). The numerical results are benchmarked against the well-established FULL code [G. Rewoldt, W. M. Tang, and M. S. Chance, Phys. Fluids 25, 480 (1982)], the PEST code [J. Manickam, Nucl. Fusion 24, 595 (1984)], and the NOVA-K code [C. Z. Cheng, Phys. Rep. 211, No. 1 (1992)]. More importantly, kinetic effects on MHD modes can be investigated nonperturbatively. In particular, the kinetic effects of the background plasma on internal kink modes and the hot particle destabilization of TAEs are studied numerically.
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
NASA Astrophysics Data System (ADS)
Sund, Richard; Scharer, John
2002-11-01
We examine a new method for generating sheared flows in advanced tokamak D-T reactors with the goal of creating and controlling internal transport barriers. Ion-Bernstein waves (IBWs) have the recognized capacity to create internal transport barriers through sheared plasma flows resulting from ion absorption. Under reactor conditions, the IBW can be generated by mode conversion of a fast magnetosonic wave incident from the high-field side (HFS) on the second harmonic resonance of a minority hydrogen component, with near 100200 MHz) minimizes parasitic absorption and permits the converted IBW to approach the fifth tritium harmonic. It also facilitates compact antennas and feeds, and efficient fast wave launch. Placement of the 5T absorption layer on the HFS is advantageous for shear production. The scheme is applicable to reactors with aspect ratio < 3 such that the conversion and absorption layers are both on the high field side of the magnetic axis. Various factors (adequate separation of the mode conversion layer from the magnetic axis, concentration of the fast wave near the midplane, large machine size, and plasma elongation) minimize poloidal field effects in the conversion zone and permit a slab analysis. We use a 1-D full-wave code to analyze the conversion and absorption. A 2-D ray-tracing code incorporating poloidal magnetic fields is used to follow the IBW for various equilibria. Within this analysis a weak bean shape appears most favorable. This is an attractive scheme for future advanced tokamak reactors. *Research supported by the Univ. of Wisconsin, Madison
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.
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.
A generalized plasma dispersion function for electron damping in tokamak plasmas
NASA Astrophysics Data System (ADS)
Berry, L. A.; Jaeger, E. F.; Phillips, C. K.; Lau, C. H.; Bertelli, N.; Green, D. L.
2016-10-01
Radio frequency wave propagation in finite temperature, magnetized plasmas exhibits a wide range of physics phenomena. The plasma response is nonlocal in space and time, and numerous modes are possible with the potential for mode conversions and transformations. In addition, diffraction effects are important due to finite wavelength and finite-size wave launchers. Multidimensional simulations are required to describe these phenomena, but even with this complexity, the fundamental plasma response is assumed to be the uniform plasma response with the assumption that the local plasma current for a Fourier mode can be described by the "Stix" conductivity. However, for plasmas with non-uniform magnetic fields, the wave vector itself is nonlocal. When resolved into components perpendicular (k⊥) and parallel (k||) to the magnetic field, locality of the parallel component can easily be violated when the wavelength is large. The impact of this inconsistency is that estimates of the wave damping can be incorrect (typically low) due to unresolved resonances. For the case of ion cyclotron damping, this issue has already been addressed by including the effect of parallel magnetic field gradients. In this case, a modified plasma response (Z function) allows resonance broadening even when k|| = 0, and this improves the convergence and accuracy of wave simulations. In this paper, we extend this formalism to include electron damping and find improved convergence and accuracy for parameters where electron damping is dominant, such as high harmonic fast wave heating in the NSTX-U tokamak, and helicon wave launch for off-axis current drive in the DIII-D tokamak.
A generalized plasma dispersion function for electron damping in tokamak plasmas
Berry, L. A.; Jaeger, E. F.; Phillips, C. K.; Lau, C. H.; Bertelli, N.; Green, D. L.
2016-10-14
Radio frequency wave propagation in finite temperature, magnetized plasmas exhibits a wide range of physics phenomena. The plasma response is nonlocal in space and time, and numerous modes are possible with the potential for mode conversions and transformations. Additionally, diffraction effects are important due to finite wavelength and finite-size wave launchers. Multidimensional simulations are required to describe these phenomena, but even with this complexity, the fundamental plasma response is assumed to be the uniform plasma response with the assumption that the local plasma current for a Fourier mode can be described by the Stix conductivity. But, for plasmas with non-uniform magnetic fields, the wave vector itself is nonlocal. When resolved into components perpendicular (k ) and parallel (k ||) to the magnetic field, locality of the parallel component can easily be violated when the wavelength is large. The impact of this inconsistency is that estimates of the wave damping can be incorrect (typically low) due to unresolved resonances. For the case of ion cyclotron damping, this issue has already been addressed by including the effect of parallel magnetic field gradients. In this case, a modified plasma response (Z function) allows resonance broadening even when k || = 0, and this improves the convergence and accuracy of wave simulations. In our paper, we extend this formalism to include electron damping and find improved convergence and accuracy for parameters where electron damping is dominant, such as high harmonic fast wave heating in the NSTX-U tokamak, and helicon wave launch for off-axis current drive in the DIII-D tokamak.
A generalized plasma dispersion function for electron damping in tokamak plasmas
Berry, L. A.; Jaeger, E. F.; Phillips, C. K.; ...
2016-10-14
Radio frequency wave propagation in finite temperature, magnetized plasmas exhibits a wide range of physics phenomena. The plasma response is nonlocal in space and time, and numerous modes are possible with the potential for mode conversions and transformations. Additionally, diffraction effects are important due to finite wavelength and finite-size wave launchers. Multidimensional simulations are required to describe these phenomena, but even with this complexity, the fundamental plasma response is assumed to be the uniform plasma response with the assumption that the local plasma current for a Fourier mode can be described by the Stix conductivity. But, for plasmas with non-uniformmore » magnetic fields, the wave vector itself is nonlocal. When resolved into components perpendicular (k ) and parallel (k ||) to the magnetic field, locality of the parallel component can easily be violated when the wavelength is large. The impact of this inconsistency is that estimates of the wave damping can be incorrect (typically low) due to unresolved resonances. For the case of ion cyclotron damping, this issue has already been addressed by including the effect of parallel magnetic field gradients. In this case, a modified plasma response (Z function) allows resonance broadening even when k || = 0, and this improves the convergence and accuracy of wave simulations. In our paper, we extend this formalism to include electron damping and find improved convergence and accuracy for parameters where electron damping is dominant, such as high harmonic fast wave heating in the NSTX-U tokamak, and helicon wave launch for off-axis current drive in the DIII-D tokamak.« less
High resolution equilibrium calculations of pedestal and SOL plasma in tokamaks
NASA Astrophysics Data System (ADS)
Medvedev, S. Yu; Martynov, A. A.; Drozdov, V. V.; Ivanov, A. A.; Poshekhonov, Yu Yu
2017-02-01
For integrated modeling of equilibrium, stability and dynamics of the divertor tokamak plasma with scrape-off layer (SOL) high resolution equilibrium calculations are needed. A new version of the CAXE equilibrium code computes the tokamak equilibrium on a numerical grid adaptive to magnetic surfaces both in the plasma region with closed flux surfaces and in the SOL region with open magnetic lines. The plasma profiles can be prescribed independently in each region with nested flux surfaces, and realistic SOL profiles with very short pressure drop off length can be accurately treated. The influence of the finite current density in SOL on the connection length is studied. From the point of view of the MHD equilibrium and stability modeling, self-consistent calculations of diverted tokamak configurations with finite current density at the separatrix require taking into account plasma outside the separatrix. Calculated high resolution equilibria provide an input to new versions of the ideal MHD stability codes treating tokamak plasma with SOL. The study of the influence of the pressure gradient profile in the pedestal plasma inside and outside the separatrix on the pedestal height limit set by external kink-ballooning mode stability is presented. Another possible application of the high resolution pedestal and SOL equilibrium code is a coupling to the SOLPS code with a purpose to increase equilibrium accuracy and support self-consistent plasma flow/equilibrium modeling.
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.; ...
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
Off-axis electron cyclotron heating and the sandpile paradigm for transport in tokamak plasmas
NASA Astrophysics Data System (ADS)
March, T. K.; Chapman, S. C.; Dendy, R. O.; Merrifield, J. A.
2004-02-01
Previous observations that suggest a substantial role for nondiffusive energy transport in tokamaks subjected to off-axis electron cyclotron heating (ECH) are compared to the output from a sandpile model. The observations considered include local and global aspects of temperature profile evolution in the DIII-D [for example, C. C. Petty and T. C. Luce, Nucl. Fusion 34, 121 (1994)] and RTP (Rijnhuizen Tokamak Project) [for example, M. R. de Baar, M. N. A. Beurskens, G. M. D. Hogeweij, and N. J. Lopes Cardozo, Phys. Plasmas 6, 4645 (1999)] tokamaks. The sandpile model employed is an extension, to incorporate noncentral fueling, of one used previously to address tokamak physics issues [S. C. Chapman, R. O. Dendy, and B. Hnat, Phys. Rev. Lett. 86, 2814 (2001)]. It is found that there are significant points of resemblance between the phenomenology of the noncentrally fueled sandpile and of the tokamaks with off-axis ECH. This suggests that the essential ingredient of the sandpile model, namely avalanching conditioned by a local critical gradient, may be one of the key transport effects generated by the tokamak plasma physics.
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.
NASA Astrophysics Data System (ADS)
Guozhong, Deng; Liang, Wang; Xiaoju, Liu; Yanmin, Duan; Jiansheng, Hu; Changzheng, Li; Ling, Zhang; Shaocheng, Liu; Huiqian, Wang; Liang, Chen; Jichan, Xu; Wei, Feng; Jianbin, Liu; Huan, Liu; Guosheng, Xu; Houyang, Guo; Xiang, Gao; the EAST Team
2017-01-01
A new pellet injection system has been equipped on the experimental advanced superconducting tokamak (EAST) in the 2012 campaign, with a pellet size of ϕ 2 mm × 2 mm, a frequency of 1 Hz-10 Hz and velocity of 150 m s-1-300 m s-1. The deuterium pellet is well-known for plasma fuelling as well as for triggering the edge localized mode (ELM). In the 2012 campaign, pellet injection experiments were successfully carried out on EAST. Temporary plasma detachment achieved by deuterium pellets has been observed in a double null (DN) divertor configuration, with multi-pellet injections at a repetition frequency of 2 Hz. The partial detachment of the outer divertors and complete detachment of the inner divertors was achieved after 35 ms of each pellet injection, which have a duration of 30-60 ms with the maximum degree of detachment (DOD) reaching 3.5 and 37, respectively. Meanwhile, the multifaceted asymmetric radiation from the edge (MARFE) phenomena was also observed at the high field side (HFS) near both the lower and upper X-points with radiation loss suddenly increased to about 15%-70%, which may be the main cause of divertor plasma detachment. The temporary detachment induced by pellet injection may act as a new way to study divertor detachment behaviors.
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.
Plasma Current Start-up by ECW and Vertical Field in the TST-2 Spherical Tokamak
NASA Astrophysics Data System (ADS)
Mitarai, Osamu; Takase, Yuichi; Ejiri, Akira; Shiraiwa, Syunichi; Kasahara, Hiroshi; Yamada, Takuma; Ohara, Shinya; TST-2 Team; Nakamura, Kazuo; Iyomasa, Atsuhiro; Hasegawa, Makoto; Idei, Hiroshi; Sakamoto, Mizuki; Hanada, Kazuaki; Satoh, Kohnosuke; Zushi, Hideki; TRIAM Group; Nishino, Nobuhiro
Plasma current start-up and ramp-up to 10 kA have been demonstrated in the TST-2 spherical tokamak without the use of the central solenoid. Only the electron cyclotron wave (ECW) and the outer equilibrium field coils are used. The plasma current evolution depends on the poloidal coil arrangement. It is also demonstrated that the plasma current start-up can take place without the field null.
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.
Electron collisionless damping of the geodesic acoustic mode in rotating tokamak plasmas
NASA Astrophysics Data System (ADS)
Xie, Baoyi; Guo, Wenfeng; Gong, Xueyu; Yu, Jun; Chen, You; Cao, Jinjia
2016-12-01
Collisionless damping of the geodesic acoustic mode due to electron dynamics in rotating tokamak plasmas is investigated. A dispersion relation of the geodesic acoustic mode with a non-adiabatic electron response in a rotating tokamak is derived and solved both analytically and numerically. It is found that the collisionless damping of the geodesic acoustic mode, due to electron dynamics, significantly increases with the increasing toroidal rotation, especially in the large safety factor regime. The rotation-induced frequency up-shift of the geodesic acoustic mode increases the resonant velocity, which enables a larger number of electrons to resonate with the geodesic acoustic mode. The significant increase of the number of the resonant electrons significantly enhances the collisionless damping of the geodesic acoustic mode. The result indicates that in rotating tokamak plasmas a more complete picture of the geodesic acoustic mode should include the electron dynamics.
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.
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.
Distinct turbulence sources and confinement features in the spherical tokamak plasma regime
Wang, W. X.; Ethier, S.; Ren, Y.; ...
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
Runaway electron mitigation by applied magnetic perturbations in RFX-mod tokamak plasmas
NASA Astrophysics Data System (ADS)
Gobbin, M.; Valisa, M.; White, R. B.; Cester, D.; Marrelli, L.; Nocente, M.; Piovesan, P.; Stevanato, L.; Puiatti, M. E.; Zuin, M.
2017-01-01
Thanks to its advanced system for the control of magnetohydrodynamic modes, the RFX-mod device run as a tokamak is particularly suited to the study of the possible impact on runaway electron (RE) de-confinement in response to applied magnetic perturbations. This paper shows that during the flat-top phase in RFX-mod discharges, with a plasma current of {{I}\\text{p}}˜ 150 kA and a low density ({{n}\\text{e}}<{{10}19} m-3), the amount of REs scales with the m = 2,n = 1 perturbation both in q(a) < 2 and q(a) > 2 plasmas. Similar results have also been obtained in post-disruption phases, but still with limited statistics. The mechanisms generating REs and the effect of magnetic perturbation (MP) on their confinement are interpreted by numerical simulations with the relativistic guiding center code ORBIT. The role played by different magnetic equilibria on the energy of REs and on their loss rates is investigated. ORBIT simulations indicate that RE-enhanced losses are associated with a raised level of stochasticity, the effect being more pronounced when the MP amplitude is higher and internally resonant.
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.
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.
A novel flexible field-aligned coordinate system for tokamak edge plasma simulation
NASA Astrophysics Data System (ADS)
Leddy, J.; Dudson, B.; Romanelli, M.; Shanahan, B.; Walkden, N.
2017-03-01
Tokamak plasmas are confined by a magnetic field that limits the particle and heat transport perpendicular to the field. Parallel to the field the ionised particles can move freely, so to obtain confinement the field lines are ;closed; (i.e. form closed surfaces of constant poloidal flux) in the core of a tokamak. Towards, the edge, however, the field lines intersect physical surfaces, leading to interaction between neutral and ionised particles, and the potential melting of the material surface. Simulation of this interaction is important for predicting the performance and lifetime of future tokamak devices such as ITER. Field-aligned coordinates are commonly used in the simulation of tokamak plasmas due to the geometry and magnetic topology of the system. However, these coordinates are limited in the geometry they allow in the poloidal plane due to orthogonality requirements. A novel 3D coordinate system is proposed herein that relaxes this constraint so that any arbitrary, smoothly varying geometry can be matched in the poloidal plane while maintaining a field-aligned coordinate. This system is implemented in BOUT++ and tested for accuracy using the method of manufactured solutions. A MAST edge cross-section is simulated using a fluid plasma model and the results show expected behaviour for density, temperature, and velocity. Finally, simulations of an isolated divertor leg are conducted with and without neutrals to demonstrate the ion-neutral interaction near the divertor plate and the corresponding beneficial decrease in plasma temperature.
Comparison Autocorrelation Method and SVD Method for Plasma Mode Analysis in Tokamaks
NASA Astrophysics Data System (ADS)
Saadat, Shervin; Salem, Mohammad K.
2011-08-01
Autocorrelation method (Single time series) is new method for analysis of plasma mode in Tokamaks. In this article autocorrelation method has been compared with SVD method. Energy of the modes which obtained by SVD analysis showed that the autocorrelation method is a cited method for mode detection.
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.
Improved Confinement in Highly Powered Advanced Tokamak Scenarios on DIII-D
NASA Astrophysics Data System (ADS)
Petrie, T. W.; Leonard, A.; Luce, T.; Osborne, T.; Solomon, W.; Turco, F.; Fenstermacher, M. E.; Holcomb, C.; Lasnier, C.; Makowski, M.
2016-10-01
DIII-D has recently demonstrated improved energy confinement by injecting neutral gas into high performance Advanced Tokamak (AT) plasmas during high power operation. Representative parameters are: q95 = 6, PIN up to 15 MW, H98 = 1.4-1.8, and βN = 2.8-4.2. Unlike in lower and moderate powered AT plasmas, τE and βN increased (and νELM decreased) as density was increased by deuterium gas puffing. We discuss how the interplay between pedestal density and temperature with fueling can lead to higher ballooning stability and a peeling/kink current limit that increasers as the pressure gradient increases. Comparison of neon, nitrogen, and argon as ``seed'' impurities in high PIN ATs in terms of their effects on core dilution, τE, and heat flux (q⊥) reduction favors argon. In general, the puff-and-pump radiating divertor was not as effective in reducing q⊥ while maintaining density control at highest PIN than it was at lower PIN. Work supported by the US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-FG02-07ER54917.
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.
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.
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.
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.
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.
Intrinsic rotation driven by non-Maxwellian equilibria in Tokamak plasmas.
Barnes, M; Parra, F I; Lee, J P; Belli, E A; Nave, M F F; White, A E
2013-08-02
The effect of small deviations from a Maxwellian equilibrium on turbulent momentum transport in tokamak plasmas is considered. These non-Maxwellian features, arising from diamagnetic effects, introduce a strong dependence of the radial flux of cocurrent toroidal angular momentum on collisionality: As the plasma goes from nearly collisionless to weakly collisional, the flux reverses direction from radially inward to outward. This indicates a collisionality-dependent transition from peaked to hollow rotation profiles, consistent with experimental observations of intrinsic rotation.
Intrinsic Rotation Driven by Non-Maxwellian Equilibria in Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Barnes, M.; Parra, F. I.; Lee, J. P.; Belli, E. A.; Nave, M. F. F.; White, A. E.
2013-08-01
The effect of small deviations from a Maxwellian equilibrium on turbulent momentum transport in tokamak plasmas is considered. These non-Maxwellian features, arising from diamagnetic effects, introduce a strong dependence of the radial flux of cocurrent toroidal angular momentum on collisionality: As the plasma goes from nearly collisionless to weakly collisional, the flux reverses direction from radially inward to outward. This indicates a collisionality-dependent transition from peaked to hollow rotation profiles, consistent with experimental observations of intrinsic rotation.
Kessel, C. E.; Poli, F. M.; Ghantous, K.; ...
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.
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.
Fusion plasma theory. Task 3: Auxiliary heating in Tokamaks and tandem mirrors
NASA Astrophysics Data System (ADS)
Scharer, J. E.
1984-06-01
The ICRF coupling, heating and breakeven studies for Tokamaks and ECRF fundamental second harmonic heating in tandem mirrors are examined. The studies have included ICRF Fokker-Planck heating and breakeven studies for large Tokamaks such as JET, fundamental work on a new wave power absorption and conservation relation for ICRF in inhomogeneous plasmas, a formulation and code development for ICRF waveguide coupling in Tokamak edge regions. The ECRF ray tracing studies were carried out for fundamental and second harmonic propagation, absorption and whistler microinstabilities in tandem mirror plug and barrier regions of Phaedrus, TMX-U and TASKA. The two-dimensional velocity space, time dependent Fokker-Planck heating studies have concentrated on D-T breakeven scenarios for fundamental minority deuterium and second harmonic tritium regimes.
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.
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.
Reconstruction of plasma current profile of tokamaks using combinatorial optimization techniques
Kishimoto, Maki; Sakasai, Kaoru; Ara, Katuyuki; Suzuki, Yasuo; Fujita, Takaaki
1996-04-01
New methods to reconstruct plasma shape and plasma current distribution from magnetic measurements are proposed. The reconstruction of plasma current profile from magnetic measurements is regarded as an optimum allocation problem of currents into cross section of the vacuum vessel of the tokamak. For solving this optimization problem, the authors use two types of solutions: a genetic algorithm and a combined method of a Hopfield neural network and a genetic algorithm. The effectiveness of these methods is shown by the application of these techniques to JT-60U plasmas.
Role of Pressure Gradient on Intrinsic Toroidal Rotation in Tokamak Plasmas
Yoshida, M.; Kamada, Y.; Takenaga, H.; Sakamoto, Y.; Urano, H.; Oyama, N.; Matsunaga, G.
2008-03-14
The toroidal plasma rotation generated by the external momentum input and by the plasma itself (intrinsic rotation) has been separated through a novel momentum transport analysis in the JT-60U tokamak device. The toroidal rotation, which is not determined by the momentum transport coefficients and the external momentum input, has been observed. It is found that this intrinsic rotation is locally determined by the local pressure gradient and increases with increasing pressure gradient. This trend is almost the same for various plasmas: low and high confinement mode, co and counterrotating plasmas.
NASA Astrophysics Data System (ADS)
Han, Hyunsun; In, Y.; Jeon, Y. M.; Lee, H. Y.; Hahn, S. H.; Lee, K. D.; Nam, Y. U.; Yoon, S. W.
2016-08-01
The change of tokamak plasma behavior by supersonic molecular beam injection (SMBI) was investigated by applying a three-dimensional magnetic perturbation that could suppress edge localized modes (ELMs). From the time trace of decreasing electron temperature and with increasing plasma density keeping the total confined energy constant, the SMBI seems to act as a cold pulse on the plasma. However, the ELM behaviors were changed drastically (i.e., the symptom of ELM suppression has disappeared). The plasma collisionality in the edge-pedestal region could play a role in the change of the ELM behaviors.
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/cm^{2}, arc-plasma cathode electron source used for local helicity injection tokamak startup. A double layer sheath is established between the high-density arc plasma (n_{arc} ≈ 10^{21} m^{-3}) within the electron source, and the less dense external tokamak edge plasma (n_{edge} ≈ 10^{18} m^{-3}) into which current is injected at the applied injector voltage, V_{inj}. Experiments on the Pegasus spherical tokamak show the injected current, I_{inj}, increases with V_{inj} according to the standard double layer scaling I_{inj} ~ V_{inj}^{3/2} at low current and transitions to I_{inj} ~ V_{inj}^{1/2} at high currents. In this high current regime, sheath expansion and/or space charge neutralization impose limits on the beam density n_{b} ~ I_{inj}/V_{inj}^{1/2}. For low tokamak edge density n_{edge} and high I_{inj}, the inferred beam density n_{b} is consistent with the requirement n_{b} ≤ n_{edge} imposed by space-charge neutralization of the beam in the tokamak edge plasma. At sufficient edge density, n_{b} ~ n_{arc} is observed, consistent with a limit to n_{b} imposed by expansion of the double layer sheath. These results suggest that n_{arc} is a viable control actuator for the source impedance.
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.
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.
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.
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.
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.
Relevant parameter space and stability of spherical tokamaks with a plasma center column
NASA Astrophysics Data System (ADS)
Lampugnani, L. G.; Garcia-Martinez, P. L.; Farengo, R.
2017-02-01
A spherical tokamak (ST) with a plasma center column (PCC) can be formed inside a simply connected chamber via driven magnetic relaxation. From a practical perspective, the ST-PCC could overcome many difficulties associated with the material center column of the standard ST reactor design. Besides, the ST-PCC concept can be regarded as an advanced helicity injected device that would enable novel experiments on the key physics of magnetic relaxation and reconnection. This is because the concept includes not only a PCC but also a coaxial helicity injector (CHI). This combination implies an improved level of flexibility in the helicity injection scheme required for the formation and sustainment phases. In this work, the parameter space determining the magnetic structure of the ST-PCC equilibria is studied under the assumption of fully relaxed plasmas. In particular, it is shown that the effect of the external bias field of the PCC and the CHI essentially depends on a single parameter that measures the relative amount of flux of these two entities. The effect of plasma elongation on the safety factor profile and the stability to the tilt mode are also analyzed. In the first part of this work, the stability of the system is explained in terms of the minimum energy principle, and relevant stability maps are constructed. While this picture provides an adequate insight into the underlying physics of the instability, it does not include the stabilizing effect of line-tying at the electrodes. In the second part, a dynamical stability analysis of the ST-PCC configurations, including the effect of line-tying, is performed by numerically solving the magnetohydrodynamic equations. A significant stability enhancement is observed when the PCC contains more than the 70% of the total external bias flux, and the elongation is not higher than two.
NASA Astrophysics Data System (ADS)
Humphreys, D. A.; Walker, M. L.; Leuer, J. A.
1999-11-01
We describe a model of linearized plasma shape and position response which is based on low poloidal mode number (m<=2, approximately vertical and major radial) displacements of the plasma current distribution. The model introduces minimal plasma degrees of freedom while providing sufficient accuracy for high performance controller design. The effects of significant variation in plasma poloidal beta, internal inductance, and separatrix configuration are taken into account. Models which can predict plasma shape and position variation with reasonable accuracy are particularly important for design of dynamic controllers in devices with significant variation in auxiliary heating input power and plasma shape --- conditions common in the DIII--D tokamak. Model predictions are validated using experimental response data from DIII--D. Application of the plasma response model to design of multivariable dynamic plasma controllers recently implemented on DIII--D is described.
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.
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
Impact of helical boundary conditions in MHD modeling of RFP and tokamak plasmas
NASA Astrophysics Data System (ADS)
Bonfiglio, D.; Cappello, S.; Escande, D. F.; Piovesan, P.; Veranda, M.; Chacón, L.
2012-10-01
Helical boundary conditions imposed by the active control system of the RFX-mod device provide a handle to govern the plasma dynamics in both RFP and Ohmic tokamak discharges [1]. By applying an edge radial magnetic field with proper helicity, it is possible to increase the persistence of the spontaneous helical RFP states at high current,and to stimulate them also at low current or high density. Helical BCs even allow to access helical states with different helicity than the spontaneous one [2]. In Ohmic tokamak operation at q(a)<2, the presence of the 2/1 RWM reduces the sawtoothing activity of the 1/1 internal kink, which takes a stationary snake-like character instead. Many of these features are qualitatively reproduced in 3D nonlinear MHD modeling. We study the impact of helical BCs on the MHD dynamics in both RFP and tokamak with two successfully benchmarked numerical tools, SpeCyl and PIXIE3D [3]. We recover the bifurcation from a sawtooth to a snake solution when imposing a 2/1 BC in the tokamak case and we interpret this as a toroidal/nonlinear coupling effect. We show that the bifurcation is more easily stimulated with a 1/1 BC.[4pt] [1] P. Piovesan, invited talk this meeting[0pt] [2] M. Veranda et al EPS-ICPP Conference (2012) P4.004[0pt] [3] D. Bonfiglio et al Phys. Plasmas (2010)
Neoclassical simulation of tokamak plasmas using the continuum gyrokinetic code TEMPEST.
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 (psi,theta,micro) 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.
ATCA digital controller hardware for vertical stabilization of plasmas in tokamaks
Batista, A. J. N.; Sousa, J.; Varandas, C. A. F.
2006-10-15
The efficient vertical stabilization (VS) of plasmas in tokamaks requires a fast reaction of the VS controller, for example, after detection of edge localized modes (ELM). For controlling the effects of very large ELMs a new digital control hardware, based on the Advanced Telecommunications Computing Architecture trade mark sign (ATCA), is being developed aiming to reduce the VS digital control loop cycle (down to an optimal value of 10 {mu}s) and improve the algorithm performance. The system has 1 ATCA trade mark sign processor module and up to 12 ATCA trade mark sign control modules, each one with 32 analog input channels (12 bit resolution), 4 analog output channels (12 bit resolution), and 8 digital input/output channels. The Aurora trade mark sign and PCI Express trade mark sign communication protocols will be used for data transport, between modules, with expected latencies below 2 {mu}s. Control algorithms are implemented on a ix86 based processor with 6 Gflops and on field programmable gate arrays with 80 GMACS, interconnected by serial gigabit links in a full mesh topology.
Multi-channel poloidal correlation reflectometry on experimental advanced superconducting tokamak
NASA Astrophysics Data System (ADS)
Qu, H.; Zhang, T.; Han, X.; Xiang, H. M.; Wen, F.; Geng, K. N.; Wang, Y. M.; Kong, D. F.; Cai, J. Q.; Huang, C. B.; Gao, Y.; Gao, X.; Zhang, S.
2016-11-01
A new multi-channel poloidal correlation reflectometry is developed at Experimental Advanced Superconducting Tokamak. Eight dielectric resonator oscillators with frequencies of 12.5 GHz, 13.5 GHz, 14.5 GHz, 15 GHz, 15.5 GHz, 16 GHz, 17 GHz, and 18 GHz are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together. The output waves are launched by one single antenna after passing through a 20 dB directional coupler which can provide the reference signal. Two poloidally separated antennae are installed to receive the reflected waves from plasma. The reference and reflected signals are down-converted by mixing with a quadrupled signal from a phase-locked source with a frequency of 14.2 GHz and the IF signals pass through the filter bank. The resulting signals from the mixers are detected by I/Q demodulators. The setup enables the measurement of density fluctuation at 8 (radial) × 2 (poloidal) spatial points. A coherent mode with an increasing velocity from 50 kHz to 100 kHz is observed by using the system. The mode is located in the steep gradient region of the pedestal.
NASA Astrophysics Data System (ADS)
Wang, H. Q.; Xu, G. S.; Guo, H. Y.; Wan, B. N.; Wang, L.; Chen, R.; Ding, S. Y.; Yan, N.; Gong, X. Z.; Liu, S. C.; Shao, L. M.; Chen, L.; Zhang, W.; Liang, Y. F.; Hu, G. H.; Liu, Y. L.; Li, Y. L.; Zhao, N.
2014-09-01
High-confinement regime with high-frequency and low-energy-loss small edge localized modes (ELMs) was achieved in Experimental Advanced Superconducting Tokamak by using the lower hybrid current drive and ion cyclotron resonance heating with lithium wall conditioning. The small ELMs are usually accompanied with a quasi-coherent mode at frequency around 30 kHz, as detected by the Langmuir probes near the separatrix. The coherent mode, with weak magnetic perturbations different from the precursor of conventional ELMs, propagates in the electron diamagnetic drift direction in the lab frame with the poloidal wavelength λθ ˜ 14 cm, corresponding to both high poloidal and toroidal mode numbers (m > 60 and n > 12). This coherent mode, carrying high-temperature high-density filament-like plasma, drives considerable transport from the pedestal region into the scrape-off layer towards divertor region. The co-existence of small ELMs and quasi-coherent modes is beneficial for the sustainment of long pulse H-mode regime without significant confinement degradation.
Multi-channel poloidal correlation reflectometry on experimental advanced superconducting tokamak.
Qu, H; Zhang, T; Han, X; Xiang, H M; Wen, F; Geng, K N; Wang, Y M; Kong, D F; Cai, J Q; Huang, C B; Gao, Y; Gao, X; Zhang, S
2016-11-01
A new multi-channel poloidal correlation reflectometry is developed at Experimental Advanced Superconducting Tokamak. Eight dielectric resonator oscillators with frequencies of 12.5 GHz, 13.5 GHz, 14.5 GHz, 15 GHz, 15.5 GHz, 16 GHz, 17 GHz, and 18 GHz are used as sources. Signals from the sources are up-converted to V band using active quadruplers and then coupled together. The output waves are launched by one single antenna after passing through a 20 dB directional coupler which can provide the reference signal. Two poloidally separated antennae are installed to receive the reflected waves from plasma. The reference and reflected signals are down-converted by mixing with a quadrupled signal from a phase-locked source with a frequency of 14.2 GHz and the IF signals pass through the filter bank. The resulting signals from the mixers are detected by I/Q demodulators. The setup enables the measurement of density fluctuation at 8 (radial) × 2 (poloidal) spatial points. A coherent mode with an increasing velocity from 50 kHz to 100 kHz is observed by using the system. The mode is located in the steep gradient region of the pedestal.
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.
Beer, M.A.; Chance, M.S.; Hahm, T.S.; Lin, Z.; Rewoldt, G.; Tang, W.M.
1997-11-01
Sheared rotation dynamics are widely believed to have signficant influence on experimentally observed confinement transitions in advanced operating modes in major tokamak experiments, such as the Tokamak Fusion Test Reactor (TFTR) [D.J. Grove and D.M. Meade, Nuclear Fusion 25, 1167 (1985)], with reversed magnetic shear regions in the plasma interior. The high-n toroidal drift modes destabilized by the combined effects of ion temperature gradients and trapped particles in toroidal geometry can be strongly affected by radially sheared toroidal and poloidal plasma rotation. In previous work with the FULL linear microinstability code, a simplified rotation model including only toroidal rotation was employed, and results were obtained. Here, a more complete rotation model, that includes contributions from toroidal and poloidal rotation and the ion pressure gradient to the total radial electric field, is used for a proper self-consistent treatment of this key problem. Relevant advanced operating mode cases for TFTR are presented. In addition, the complementary problem of the dynamics of fluctuation-driven E x B flow is investigated by an integrated program of gyrokinetic simulation in annulus geometry and gyrofluid simulation in flux tube geometry.
Rewoldt, G.; Beer, M.A.; Chance, M.S.; Hahm, T.S.; Lin, Z.; Tang, W.M.
1997-12-01
Sheared rotation dynamics are widely believed to have significant influence on experimentally observed confinement transitions in advanced operating modes in major tokamak experiments, such as the Tokamak Fusion Test Reactor (TFTR) with reversed magnetic shear regions in the plasma interior. The high-n toroidal drift modes destabilized by the combined effects of ion temperature gradients and trapped particles in toroidal geometry can be strongly affected by radially sheared toroidal and poloidal plasma rotation. In previous work with the FULL linear microinstability code, a simplified rotation model including only toroidal rotation was employed, and results were obtained. Here, a more complete rotation model, that includes contributions from toroidal and poloidal rotation and the ion pressure gradient to the total radial electric field, is used for a proper self-consistent treatment of this key problem. Relevant advanced operating mode cases for TFTR are presented. In addition, the complementary problem of the dynamics of fluctuation-driven E x B flow is investigated by an integrated program of gyrokinetic simulation in annulus geometry and gyrofluid simulation in flux tube geometry.
Plasma flows and fluctuations with magnetic islands in the edge plasmas of J-TEXT tokamak
NASA Astrophysics Data System (ADS)
Zhao, K. J.; Shi, Y. J.; Hahn, S. H.; Diamond, P. H.; Sun, Y.; Cheng, J.; Liu, H.; Lie, N.; Chen, Z. P.; Ding, Y. H.; Chen, Z. Y.; Rao, B.; Leconte, M.; Bak, J. G.; Cheng, Z. F.; Gao, L.; Zhang, X. Q.; Yang, Z. J.; Wang, N. C.; Wang, L.; Jin, W.; Yan, L. W.; Dong, J. Q.; Zhuang, G.; J-TEXT Team
2015-07-01
The first comprehensive measurements of plasma flows and fluctuations nearby static magnetic islands driven by resonant magnetic perturbations are presented. These experiments were performed using multiple Langmuir probe arrays on the edge plasmas of the J-TEXT tokamak. Controlled variations of the island size and location are explored. This study aims to understand the interaction between turbulence and magnetic islands, and to elucidate magnetic island effects on edge turbulence and flow intensity profiles, edge electric fields, and thus confinement regime transitions. Turbulence and low frequency flows (LFFs) all drop inside the magnetic island, but increase at its boundary, as island width increases. The geodesic acoustic mode is damped in most of the edge area with magnetic islands. The sign of the radial electric field changes from negative to positive within the islands. The gradient of turbulent stresses vanishes at the island center, and becomes steeper at the boundaries of the islands. The particle transport induced by the turbulence is reduced inside the magnetic islands. The magnetic island effects on flows and turbulence can lead to an increase in LFFs and enhance Reynolds stresses near the last closed flux surface (LCFS). A stronger radial electric field layer can be formed near the LCFS when magnetic islands are present. The results suggest that magnetic islands can be used as a tool to enhance edge turbulence and flows, edge electric fields, and thus to trigger confinement regime transitions.
Ehst, D.A.; Hassanein, A.
1996-02-01
Ablation damage to solid targets with high heat flux impulses is generally greater high-energy electron beam heat sources compared to low-energy plasma guns. This sensitivity to incoming particle kinetic energy is explored with computer modelling; a fast-running routine (DESIRE) is developed for initial scoping analysis and is found to be in reasonable agreement with several experiments on graphite and tungsten targets. If tokamak disruptions are characterized by particle energies less than {approximately}1 keV, then we expect plasma guns are a better analogue than electron beams for simulating disruption behavior and testing candidate plasma-facing materials.
NASA Astrophysics Data System (ADS)
Han, Hyunsun; in, Y.; Jeon, Y. M.; Hahn, S. H.; Lee, K. D.; Nam, Y. U.; Yoon, S. W.
2016-10-01
In KSTAR experiments, the change of tokamak plasma behavior by supersonic molecular beam injection (SMBI) was investigated by applying resonant magnetic perturbations(RMP) that could suppress edge localized modes (ELMs). When the SMBI is applied, the symptom representing ELM suppression by RMP is disappeared. The SMBI acts as a cold pulse on the plasma keeping the total confinement engergy constant. However, it makes plasma density increase and change the plasama collisionality which can play a role in the edge-pedestal build-up processing. This work was supported by Project PG1201-2 and the KSTAR research project funded by Korea Ministry of Science, ICT and Future Planning.
Advancing the understanding of plasma transport in mid-size stellarators
NASA Astrophysics Data System (ADS)
Hidalgo, Carlos; Talmadge, Joseph; Ramisch, Mirko; TJ-II, the; HXS; TJ-K Teams
2017-01-01
The tokamak and the stellarator are the two main candidate concepts for magnetically confining fusion plasmas. The flexibility of the mid-size stellarator devices together with their unique diagnostic capabilities make them ideally suited to study the relation between magnetic topology, electric fields and transport. This paper addresses advances in the understanding of plasma transport in mid-size stellarators with an emphasis on the physics of flows, transport control, impurity and particle transport and fast particles. The results described here emphasize an improved physics understanding of phenomena in stellarators that complements the empirical approach. Experiments in mid-size stellarators support the development of advanced plasma scenarios in Wendelstein 7-X (W7-X) and, in concert with better physics understanding in tokamaks, may ultimately lead to an advance in the prediction of burning plasma behaviour.
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.
Impact of lithium pellets on plasma performance in the ASDEX Upgrade all-metal-wall tokamak
NASA Astrophysics Data System (ADS)
Lang, P. T.; Maingi, R.; Mansfield, D. K.; McDermott, R. M.; Neu, R.; Wolfrum, E.; Arredondo Parra, R.; Bernert, M.; Birkenmeier, G.; Diallo, A.; Dunne, M.; Fable, E.; Fischer, R.; Geiger, B.; Hakola, A.; Nikolaeva, V.; Kappatou, A.; Laggner, F.; Oberkofler, M.; Ploeckl, B.; Potzel, S.; Pütterich, T.; Sieglin, B.; Szepesi, T.; ASDEX Upgrade Team
2017-01-01
The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6 × 1020 Li atoms were launched with a speed of 600 m s-1 to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%.
Internal transport barrier simulation with pellet injection in tokamak and helical reactor plasmas
NASA Astrophysics Data System (ADS)
Higashiyama, Y.; Yamazaki, K.; Garcia, J.; Arimoto, H.; Shoji, T.
2008-07-01
In the future fusion reactor, the plasma density peaking is important for the increase in the fusion power gain. The density control and the internal transport barrier (ITB) formation due to the pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. Firstly, the pellet injection simulation is carried out including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of the high field side (HFS) pellet injection is clarified. Secondly, the ITB simulation with the pellet injection is carried out with the confinement improvement model based on the E×B shear effects, and it is found that the deep pellet penetration is helpful for the ITB formation as well as the plasma core fuelling in the reversed shear tokamak reactor, but the deep pellet penetration is not effective in the helical reactor.
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.
Plasma Core Electron Density and Temperature Measurements Using CVI Line Emissions in TCABR Tokamak
NASA Astrophysics Data System (ADS)
do Nascimento, F.; Machida, M.; Severo, J. H. F.; Sanada, E.; Ronchi, G.
2015-08-01
In this work, we present results of electron temperature ( T e ) and density ( n e ) measurements obtained in Tokamak Chauffage Alfvén Brésilien (TCABR) tokamak using visible spectroscopy from CVI line emissions which occurs mainly near the center of the plasma column. The presented method is based on a well-known relationship between the particle flux ( Γ ion) and the photon flux ( ø ion) emitted by an ion species combined with ionizations per photon atomic data provided by the atomic data and analysis structure (ADAS) database. In the experiment, we measured the photon fluxes of three different CVI spectral line emissions, 4685.2, 5290.5, and 6200.6 Å (one line per shot). Using this method it was possible to find out the temporal evolution of T e and n e in the plasma. The results achieved are in good agreement with T e and n e measurements made using other diagnostic tools.
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.
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.
Technology Advances in Support of Fusion Plasma Imaging Diagnostics
NASA Astrophysics Data System (ADS)
Jiang, Qi; Lai, Jiali; Hu, Fengqi; Li, Maijou; Chang, Yu-Ting; Domier, Calvin; Luhmann, Neville, Jr.
2012-10-01
Innovative technologies are under investigation in key areas to enhance the performance of microwave and millimeter-wave fusion plasma imaging diagnostics. Novel antenna and mixer configurations are being developed at increasingly higher frequencies, to facilitate the use of electron cyclotron emission imaging (ECEI) on high field (> 2.6 T) plasma devices. Low noise preamplifier-based imaging antenna arrays are being developed to increase the sensitivity and dynamic range of microwave imaging reflectometry (MIR) diagnostics for the localized measurement of turbulent density fluctuations. High power multi-frequency sources, fabricated using advanced CMOS technology, offer the promise of allowing MIR-based diagnostic instruments to image these density fluctuations in 2-D over an extended plasma volume in high performance tokamak plasmas. Details regarding each of these diagnostic development areas will be presented.
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.
The construction of an electrode biasing system for driving plasma rotation in J-TEXT tokamak
NASA Astrophysics Data System (ADS)
Zhu, T. Z.; Chen, Z. P.; Sun, Yue; Nan, J. Y.; Liu, H.; Zhuang, G.; Wang, Z. J.
2014-05-01
A newly designed electrode biasing system has been constructed for driving plasma rotation in J-TEXT tokamak. To reduce the influence to the plasma, the system contains a pneumatic driving system so that it can reciprocate in a single discharge, with a stroke of about 5 cm in 100 ms. The power supply of the system can provide stable and adjustable dc voltage in the range of 0-700 V, with adjustable duration of 10-200 ms; its instantaneous power output can reach up to more than 200 kW. In addition, the power supply can also provide a multi-cycle voltage waveform, with adjustable pulse width and voltage amplitude. When applying a positive bias to the plasma, both an improvement of plasma confinement and the speed-up of plasma-edge toroidal rotation in the same direction of plasma current are observed in the experiments.
The construction of an electrode biasing system for driving plasma rotation in J-TEXT tokamak.
Zhu, T Z; Chen, Z P; Sun, Yue; Nan, J Y; Liu, H; Zhuang, G; Wang, Z J
2014-05-01
A newly designed electrode biasing system has been constructed for driving plasma rotation in J-TEXT tokamak. To reduce the influence to the plasma, the system contains a pneumatic driving system so that it can reciprocate in a single discharge, with a stroke of about 5 cm in 100 ms. The power supply of the system can provide stable and adjustable dc voltage in the range of 0-700 V, with adjustable duration of 10-200 ms; its instantaneous power output can reach up to more than 200 kW. In addition, the power supply can also provide a multi-cycle voltage waveform, with adjustable pulse width and voltage amplitude. When applying a positive bias to the plasma, both an improvement of plasma confinement and the speed-up of plasma-edge toroidal rotation in the same direction of plasma current are observed in the experiments.
On the mechanisms governing gas penetration into a tokamak plasma during a massive gas injection
NASA Astrophysics Data System (ADS)
Nardon, E.; Fil, A.; Chauveau, P.; Tamain, P.; Guirlet, R.; Koslowski, H. R.; Lehnen, M.; Reux, C.; Saint-Laurent, F.; Contributors, JET
2017-01-01
A new 1D radial fluid code, IMAGINE, is used to simulate the penetration of gas into a tokamak plasma during a massive gas injection (MGI). The main result is that the gas is in general strongly braked as it reaches the plasma, due to mechanisms related to charge exchange and (to a smaller extent) recombination. As a result, only a fraction of the gas penetrates into the plasma. Also, a shock wave is created in the gas which propagates away from the plasma, braking and compressing the incoming gas. Simulation results are quantitatively consistent, at least in terms of orders of magnitude, with experimental data for a D 2 MGI into a JET Ohmic plasma. Simulations of MGI into the background plasma surrounding a runaway electron beam show that if the background electron density is too high, the gas may not penetrate, suggesting a possible explanation for the recent results of Reux et al in JET (2015 Nucl. Fusion 55 093013).
The Reconstruction of the Plasma Boundary in the Sino-United Spherical Tokamak Experiments
NASA Astrophysics Data System (ADS)
Liu, Jian; Feng, Chunhua; Yang, Xuanzong; Wang, Long
2010-04-01
A method for the reconstruction of the plasma boundary in the sino-united spherical tokamak (SUNIST) based on the outer plasma magnetic diagnostics is reported. In SUNIST, the magnetic flux loop integral signals were measured recently and the plasma boundary could be reconstructed well with a current filament (CF) model by setting 2 to 8 current filaments. There are three additional filament positional parameters in addition to the filament current to minimize the square root error in the CF model. The plasma configuration obtained with the CF method is consistent with the visible plasma image from the CCD camera. The average difference in the minor radii for the plasma boundary, by applying the CF model and EFIT code, is below 6 mm.
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.
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.
Calculations of axisymmetric stability of tokamak plasmas with active and passive feedback
Ward, D.J.; Jardin, S.C.; Cheng, C.Z.
1991-07-01
A new linear MHD stability code, NOVA-W, has been developed in order to study feedback stabilization of the axisymmetric mode in deformable tokamak plasmas. The NOVA-W code is a modification of the non-variational MHD stability code NOVA that includes the effects of resistive passive conductors and active feedback circuits. The vacuum calculation has been reformulated in terms of the perturbed poloidal flux to allow the inclusion of perturbed toroidal currents outside the plasma. The boundary condition at the plasma-vacuum interface relates the instability displacement to the perturbed poloidal flux. This allows a solution of the linear MHD stability equations with the feedback effects included. The passive stability predictions of the code have been tested both against a simplified analytic model and against a different numerical calculation for a realistic tokamak configuration. The comparisons demonstrate the accuracy of the NOVA-W results. Active feedback calculations are performed for the CIT tokamak design demonstrating the effect of varying the position of the flux loops that provide the measurements of vertical displacement. The results compare well with those computed earlier using a less efficient nonlinear code. 37 refs., 13 figs.
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.
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.
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.
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.
Remote operation of the GOLEM tokamak with hydrogen and helium plasmas
NASA Astrophysics Data System (ADS)
Svoboda, V.; Dvornova, A.; Dejarnac, R.; Prochazka, M.; Zaprianov, S.; Akhmethanov, R.; Bogdanova, M.; Dimitrova, M.; Dimitrov, Zh; Grover, O.; Hlavata, L.; Ivanov, K.; Kruglov, K.; Marinova, P.; Masherov, P.; Mogulkin, A.; Mlynar, J.; Stockel, J.; Volynets, A.
2016-10-01
The GOLEM tokamak was operated remotely via Internet connection during the 6th International Workshop and Summer School on Plasma Physics. Performances of hydrogen and helium discharges are compared in this paper. It is found, at similar vacuum conditions, that helium discharges are shorter but the breakdown of the working gas can be quite easily achieved at almost the same loop voltage. The plasma current in helium discharges is slightly lower than in the case of hydrogen. Turbulent fluctuations of the floating potential measured by means of an array of Langmuir probes reveal a noticeably different character in the two discharges.
Angioni, C.; Casson, F. J.; Veth, C.; Peeters, A. G.
2012-12-15
A local gyrokinetic description of the centrifugal effects on impurity transport in tokamak plasmas is presented, which extends previous models with the inclusion of the gradient of the background toroidal angular velocity in the equilibrium distribution. The equations are implemented in a gyrokinetic code. An analytical model is derived and formulae are proposed which allow the calculation of centrifugal effects on impurity transport in the limit where centrifugal effects are large only for heavy impurities in trace concentration, but are negligible for the bulk plasma. The analytic formulae are shown to be in quantitative agreement with the numerical results and are proposed for complementing present transport models.
Edge temperature gradient as intrinsic rotation drive in Alcator C-Mod tokamak plasmas.
Rice, J E; Hughes, J W; Diamond, P H; Kosuga, Y; Podpaly, Y A; Reinke, M L; Greenwald, M J; Gürcan, Ö D; Hahm, T S; Hubbard, A E; Marmar, E S; McDevitt, C J; Whyte, D G
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 ∇T, but completely different edge ∇n, it may be concluded that the drive of the intrinsic rotation is the edge ∇T rather than ∇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.
Sawtooth Pacing by Real-Time Auxiliary Power Control in a Tokamak Plasma
Goodman, T. P.; Felici, F.; Sauter, O.; Graves, J. P.
2011-06-17
In the standard scenario of tokamak plasma operation, sawtooth crashes are the main perturbations that can trigger performance-degrading, and potentially disruption-generating, neoclassical tearing modes. This Letter demonstrates sawtooth pacing by real-time control of the auxiliary power. It is shown that the sawtooth crash takes place in a reproducible manner shortly after the removal of that power, and this can be used to precisely prescribe, i.e., pace, the individual sawteeth. In combination with preemptive stabilization of the neoclassical tearing modes, sawtooth pacing provides a new sawtooth control paradigm for improved performance in burning plasmas.
Final Report: Spectral Analysis of L-shell Data in the Extreme Ultraviolet from Tokamak Plasmas
Lepson, J.; Jernigan, J. Garrett; Beiersdorfer, P.
2016-02-05
We performed detailed analyses of extreme ultraviolet spectra taken by Lawrence Livermore National Laboratory on the National Spherical Torus Experiment at Princeton Plasma Physics Laboratory and on the Alcator CKmod tokamak at the M.I.T. Plasma Science and Fusion Center. We focused on the emission of iron, carbon, and other elements in several spectral band pass regions covered by the Atmospheric Imaging Assembly on the Solar Dynamics Observatory. We documented emission lines of carbon not found in currently used solar databases and demonstrated that this emission was due to charge exchange.
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 equilibrium and confinement in a tokamak with nearly zero central current density in JT-60U.
Fujita, T; Oikawa, T; Suzuki, T; Ide, S; Sakamoto, Y; Koide, Y; Hatae, T; Naito, O; Isayama, A; Hayashi, N; Shirai, H
2001-12-10
A high confinement equilibrium with nearly zero toroidal current in the central region (a "current hole") has been observed for the first time to persist stably for several seconds in the JT-60U tokamak. This observation indicates the possibility of stable tokamak operation without central toroidal current; the central current has previously been believed to be necessary in tokamaks. The radius of the current hole extended up to 40% of the plasma minor radius. It was observed that the current hole was formed by the increase of the off-axis noninductive current.
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.
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.
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. 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.
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.
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.
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.
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.
Effects of Equilibrium Toroidal Flow on Locked Mode and Plasma Response in a Tokamak
NASA Astrophysics Data System (ADS)
Zhu, Ping; Huang, Wenlong; Yan, Xingting
2016-10-01
It is widely believed that plasma flow plays significant roles in regulating the processes of mode locking and plasma response in a tokamak in presence of external resonant magnetic perturbations (RMPs). Recently a common analytic relation for both locked mode and plasma response has been developed based on the steady-state solution to the coupled dynamic system of magnetic island evolution and torque balance. The analytic relation predicts the size of the magnetic island of a locked mode or a static nonlinear plasma response for a given RMP amplitude, and rigorously proves a screening effect of the equilibrium toroidal flow. To test the theory, we solve for the locked mode and the nonlinear plasma response in presence of RMP for a circular-shaped limiter tokamak equilibrium with constant toroidal flow, using the initial-value, full MHD simulation code NIMROD. The comparison between the simulation results and the theory prediction, in terms of the quantitative screening effects of equilibrium toroidal flow, will be reported and discussed. Supported by National Magnetic Confinement Fusion Science Program of China Grants 2014GB124002 and 2015GB101004, the 100 Talent Program of the Chinese Academy of Sciences, and U.S. Department of Energy Grants DE-FG02-86ER53218 and DE-FC02-08ER54975.
Suppression of runaway electrons with a resonant magnetic perturbation in MST tokamak plasmas
NASA Astrophysics Data System (ADS)
Munaretto, Stefano; Chapman, B. E.; Almagri, A. F.; Cornille, B. S.; Dubois, A. M.; Goetz, J. A.; McCollam, K. J.; Sovinec, C. R.
2016-10-01
Runaway electrons generated in MST tokamak plasmas are now being probed with resonant magnetic perturbations (RMP's). An RMP with m =3 strongly suppresses the runaway electrons. Initial modeling of these plasmas with NIMROD shows the degradation of flux surfaces with an m =3 RMP, which may account for the runaway electron suppression. These MST tokamak plasmas have Bt =0.14 T, Ip =50kA, and q(a) =2.2, with a bulk electron density and temperature of 5x1017 m-3 and 150 eV. Runaway electrons are detected via x-ray emission. The RMP is produced by a poloidal array of 32 saddle coils at the narrow vertical insulated cut in MST's thick conducting shell. Each RMP has a single m but a broad n spectrum. A sufficiently strong m =3 RMP completely suppresses the runaway electrons, while a comparable m =1 RMP has little effect. The impact of the RMP's on the magnetic topology of these plasmas is being studied with the nonlinear MHD code, NIMROD. With an m =3 RMP, stochasticity is introduced in the outer third of the plasma. No such change is observed with the m =1 RMP. NIMROD also predicts regularly occurring sawtooth oscillations with a period comparable to MHD activity observed in the experiment. Work supported by USDOE.
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.
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.
Sawtooth stabilization by localized electron cyclotron heating in a tokamak plasma
Hanada, K.; Tanaka, H.; Iida, M.; Minami, T.; Maekawa, T.; Terumichi, Y.; Tanaka, S. . Dept. of Physics); Ide, S. . Naka Fusion Research Establishment); Nakamura, M. ); Yamada, M.; Manickam, J.; White, R.B. . Plasma Physics Lab.)
1990-11-01
Sawtooth oscillations (STO) in the ohmically heated WT-3 tokamak are strongly modified or suppressed by localized electron cyclotron resonance heating (ECH) near the q = 1 surface, where q refers to the safety factor. The effect of ECH is much stronger when it is applied on the high field side (the inner side of the tokamak) as compared to the low field side (outer side). Complete suppression of the STO is achieved for the duration of the ECH when it is applied on the high field side, in a low density plasma, provided the ECH power exceeds a thresholds value. The STO stabilization is attributed to a modification of the current density profile by hot electrons generated by ECH, which reduces the shear in the q = region. 14 refs., 5 figs.
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.
Compatibility of the Zr-Al alloy with a tokamak plasma environment
Knize, R.J.; Cecchi, J.L.; Dylla, H.F.
1981-12-01
We have investigated the compatibility of the Zr-Al alloy bulk getter with a tokamak plasma environment, where the hydrogenic fluxes are sufficient to cause embrittlement in relatively short times. Under normal operating conditions with the getters activated, it is necessary to regenerate the absorbed hydrogenic species before the embrittlement limit is reached. We present a method for determining the loading under tokamak conditions where the Zr-Al surface characteristics can change. During glow discharge cleaning and pulse discharge cleaning, it is not convenient to regenerate. We find, however, that during the cleaning operations the getter self-inerts, thus limiting the loading. We present data and a model which accounts for this behavior in terms of impurity adsorption on the room temperature getter surface during the cleaning operations.
NASA Astrophysics Data System (ADS)
Lyublinski, I. E.; Vertkov, A. V.; Zharkov, M. Yu; Sevryukov, O. N.; Dzhumaev, P. S.; Shumskiy, V. A.; Ivannikov, A. A.
2016-09-01
Capillary-Pore Systems (CPS) filled by liquid metals are considered as an alternative solution of materials choice for plasma facing component of tokamak reactor. Tin is viewed as one of the candidates for CPS because it has lower corrosiveness than gallium and lower saturated vapour pressure compared to lithium. The corrosion resistance of Mo, Nb and W in pure liquid tin was investigated. The corrosion tests were carried out in the static isothermal conditions at a temperature up to 1050°C. As a result of the corrosion study, it was found that Mo does not corrode in liquid Sn, as opposed to Nb and is compatible with liquid tin in temperatures of up to approx. 1000°C. This allows considering Mo as an alloy base material of the in-vessel tokamak elements based on liquid tin capillary pore systems.
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.
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.
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.
On Stochastic Control of Tokamak and Artificial Intelligence
NASA Astrophysics Data System (ADS)
Rastovic, Danilo
2007-12-01
Instead of the theory of invariant manifolds and entropy reduction, the theory of fractional Brownian motions and artificiall neural networks is used for description of advanced methods for control of tokamak plasma behaviour.
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.
Interaction of neutral atoms and plasma turbulence in the tokamak edge region
NASA Astrophysics Data System (ADS)
Wersal, Christoph; Ricci, Paolo; Jorge, Rogerio; Morales, Jorge; Paruta, Paola; Riva, Fabio
2016-10-01
A novel first-principles self-consistent model that couples plasma and neutral atom physics suitable for the simulation of turbulent plasma behaviour in the tokamak edge region has been developed and implemented in the GBS code. While the plasma is modelled by the drift-reduced two fluid Braginskii equations, a kinetic model is used for the neutrals, valid in short and in long mean free path scenarios. The model includes ionization, charge-exchange, recombination, and elastic collisional processes. The model was used to study the transition form the sheath to the conduction limited regime, to include gas puffs in the simulations, and to investigate the interplay between neutral atoms and plasma turbulence.
Chang, C S; Ku, Seung-Hoe; Diamond, P. H.; Adams, Mark; Tchoua, Roselyne B; Chen, Yang; Cummings, J.; D'Azevedo, Ed F; Dif-Pradalier, Guilhem; Ethier, Stephane; Greengard, Leslie; Hahm, Taik Soo; Hinton, Fred; Keyes, David E; Klasky, Scott A; Lin, Z.; Lofstead, J.; Park, G.; Podhorszki, Norbert; Schwan, Karsten; Shoshani, A.; Silver, D.; Wolf, M.; Worley, Patrick H; Zorin, Denis
2009-01-01
Performance prediction for ITER is based upon the ubiquitous experimental observation that the plasma energy confinement in the device core is strongly coupled to the edge confinement for an unknown reason. The coupling time-scale is much shorter than the plasma transport time-scale. In order to understand this critical observation, a multi-scale turbulence-neoclassical simulation of integrated edge-core plasma in a realistic diverted geometry is a necessity, but has been a formidable task. Thanks to the recent development in high performance computing, we have succeeded in the integrated multiscale gyrokinetic simulation of the ion-temperature-gradient driven turbulence in realistic diverted tokamak geometry for the first time. It is found that modification of the self-organized criticality in the core plasma by nonlocal core-edge coupling of ITG turbulence can be responsible for the core-edge confinement coupling.
Chang, C S; Ku, Seung-Hoe; Diamond, Patrick; Adams, Mark; Tchoua, Roselyne B; Chen, Yang; Cummings, Julian; D'Azevedo, Eduardo; Dif-Pradalier, Guilhem; Ethier, Stephane; Greengard, Leslie; Hahm, Taik Soo; Hinton, Fred; Keyes, David E; Klasky, Scott A; Lin, Zhihong; Lofstead, J.; Park, G.; Parker, Scott; Podhorszki, Norbert; Schwan, Karsten; Shoshani, A.; Silver, D.; Weitzner, Harold; Wolf, M.; Worley, Patrick H; Yoon, E.; Zorin, Denis
2009-01-01
Performance prediction for ITER is based upon the ubiquitous experimental observation that the plasma energy confinement in the device core is strongly coupled to the edge confinement for an unknown reason. The coupling time-scale is much shorter than the plasma transport time-scale. In order to understand this critical observation, a multi-scale turbulence-neoclassical simulation of integrated edge-core plasma in a realistic diverted geometry is a necessity, but has been a formidable task. Thanks to the recent development in high performance computing, we have succeeded in the integrated multiscale gyrokinetic simulation of the ion-temperature-gradient driven turbulence in realistic diverted tokamak geometry for the first time. It is found that modification of the self-organized criticality in the core plasma by nonlocal core-edge coupling of ITG turbulence can be responsible for the core-edge confinement coupling.
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.
NASA Astrophysics Data System (ADS)
Ding, B. J.; Li, M. H.; Li, Y. C.; Wang, M.; Liu, F. K.; Shan, J. F.; Li, J. G.; Wan, B. N.; Wan
2017-02-01
Aiming at a fusion reactor, two issues must be solved for the lower hybrid current drive (LHCD), namely good lower hybrid wave (LHW)-plasma coupling and effective current drive at high density. For this goal, efforts have been made to improve LHW-plasma coupling and current drive capability at high density in experimental advanced superconducting tokamak (EAST). LHW-plasma coupling is improved by means of local gas puffing and gas puffing from the electron side is taken as a routine way for EAST to operate with LHCD. Studies of high density experiments suggest that low recycling and high lower hybrid (LH) frequency are preferred for LHCD experiments at high density, consistent with previous results in other machines. With the combination of 2.45 GHz and 4.6 GHz LH waves, a repeatable high confinement mode plasma with maximum density up to 19~\\text{m}-3$ was obtained by LHCD in EAST. In addition, in the first stage of LHCD cyclic operation, an alternative candidate for more economical fusion reactors has been demonstrated in EAST and further work will be continued.
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.
NASA Astrophysics Data System (ADS)
Nieto, M.; Allain, J. P.; Hassanein, A.; Titov, V.; Hendricks, M.; Gray, T.; Kaita, R.; Kugel, H.; Majeski, R.; Mansfield, D.; Spaleta, J.; Timberlake, J.
2006-12-01
The role of lithium on the modification of recycling regimes in fusion reactors has renewed interest of previous lithium supershot experiments carried out in TFTR. There is a need to understand the interaction between edge plasmas and lithiated plasma-facing components (PFCs), which have the potential of enabling fusion reactors to operate at low-recycling regimes. The Interaction of Materials with Particles and Components Testing (IMPACT) facility at Argonne National Laboratory is currently collaborating with Princeton Plasma Physics Laboratory (PPPL) to conduct lithiated surface studies for the National Spherical Tokamak Experiment (NSTX) and the Current Drive eXperiment — Upgrade (CDX-U). IMPACT has the necessary tools to perform experiments that diagnose the surface dynamics of lithium thin films on metallic and non-metallic substrates, and can be monitored with multiple in-situ techniques (LEISS, AES, QMS and XPS) capturing real-time surface dynamics. Therefore, these techniques are available during He+ and D+ irradiation. Surface sputtering measurements can be performed using a quartz crystal microbalance — dual crystal unit (QCM-DCU) with very high sensitivity. Initial results suggest that lithium intercalation into graphite occurs quite rapidly and only a fraction lithium can be kept on the surface. On metallic substrates this intercalation is absent. Additional results of Li/metal systems show lithium surface self-healing with temperature. It was also found that the presence of lithium seems to inhibit hydrocarbon formation during D+ bombardment of graphite. Experiments in CDX-U have tested the effect of both solid and liquid lithium PFCs on tokamak plasmas, and significant changes in tokamak operation are observed. These include a strong reduction in both recycling and impurity levels in the gas phase, lowered loop voltage during ohmic operation, and an increased electron temperature at the edge.
Modeling of transport phenomena in tokamak plasmas with neural networks
NASA Astrophysics Data System (ADS)
Meneghini, O.; Luna, C. J.; Smith, S. P.; Lao, L. L.
2014-06-01
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.
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.
Shaping Effects on Resistive-Plasma Resistive-Wall Mode Stability in a Tokamak
NASA Astrophysics Data System (ADS)
Rhodes, Dov; Cole, A. J.; Navratil, G. A.; Levesque, J. P.; Mauel, M. E.; Brennan, D. P.; Finn, J. M.; Fitzpatrick, R.
2016-10-01
A sharp-boundary MHD model is used to explore the effects of toroidal curvature and cross-sectional shaping on resistive-plasma resistive-wall modes in a tokamak. Building on the work of Fitzpatrick, we investigate mode stability with fixed toroidal number n =1 and a broad spectrum of poloidal m-numbers, given varying aspect-ratio, elongation, triangularity and up-down asymmetry. The speed and versatility of the sharp-boundary model facilitate exploration of a large parameter space, revealing qualitative trends to be further investigated by larger codes. In addition, the study addresses the effect of geometric mode-coupling on higher beta stability limits associated with an ideal-plasma or ideal-wall. These beta limits were used by Brennan and Finn to identify plasma response domains for feedback control. Present results show how geometric mode-coupling affects the stability limits and plasma response domains. The results are explained by an analytic reduced-MHD model with two coupled modes having different m-numbers. The next phase of this work will explore feedback control in different tokamak geometries. Supported by U.S. DOE Grant DE-FG02-86ER53222.
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.
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.
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.
Modeling the effect of toroidal plasma rotation on drift-magnetohydrodynamic modes in tokamaks
NASA Astrophysics Data System (ADS)
Chapman, I. T.; Sharapov, S. E.; Huysmans, G. T. A.; Mikhailovskii, A. B.
2006-06-01
A new code, MISHKA-F (Flow), has been developed as an extension of the ideal magneto-hydrodynamic (MHD) code MISHKA-1 [Mikhailovskii et al., Plasma Phys. Rep. 23, 844 (1997)] in order to investigate the linear MHD stability of ideal and resistive eigenmodes with respect to the effects of toroidal rotation in tokamaks in general toroidal geometry with the ion diamagnetic drift effect taken into account. Benchmark test results of the MISHKA-F code show good agreement with analytic theory [A. B. Mikhailovskii and S. E. Sharapov, Plasma Phys. Controlled Fusion 42, 57 (2000)] for the stability limits of the ideal n /m=1/1 internal kink mode. The combined stabilizing effects of the ion diamagnetic drift frequency, ω*i, and the toroidal flow shear are also studied. The ω*i stabilization of the internal kink mode is found to be more effective at finite flow shear. Finite-n ballooning modes are studied in plasmas with the toroidal flow shear effect included. The stabilization of the ballooning modes by toroidal rotation is found to agree well with earlier predictions [Webster et al., Phys. Plasmas 11, 2135 (2004)]. The effect of high flow shear is analyzed for a sawtoothing discharge typical in the Mega Ampère Spherical Tokamak (MAST) [Sykes et al., Nucl. Fusion 41, 1423 (2001)]. It is found that the ideal n =1 internal kink mode can be stabilized by toroidal rotation at values observed experimentally.
Modeling the effect of toroidal plasma rotation on drift-magnetohydrodynamic modes in tokamaks
Chapman, I.T.; Sharapov, S.E.; Huysmans, G.T.A.; Mikhailovskii, A. B.
2006-06-15
A new code, MISHKA-F (Flow), has been developed as an extension of the ideal magneto-hydrodynamic (MHD) code MISHKA-1 [Mikhailovskii et al., Plasma Phys. Rep. 23, 844 (1997)] in order to investigate the linear MHD stability of ideal and resistive eigenmodes with respect to the effects of toroidal rotation in tokamaks in general toroidal geometry with the ion diamagnetic drift effect taken into account. Benchmark test results of the MISHKA-F code show good agreement with analytic theory [A. B. Mikhailovskii and S. E. Sharapov, Plasma Phys. Controlled Fusion 42, 57 (2000)] for the stability limits of the ideal n/m=1/1 internal kink mode. The combined stabilizing effects of the ion diamagnetic drift frequency, {omega}{sub *i}, and the toroidal flow shear are also studied. The {omega}{sub *i} stabilization of the internal kink mode is found to be more effective at finite flow shear. Finite-n ballooning modes are studied in plasmas with the toroidal flow shear effect included. The stabilization of the ballooning modes by toroidal rotation is found to agree well with earlier predictions [Webster et al., Phys. Plasmas 11, 2135 (2004)]. The effect of high flow shear is analyzed for a sawtoothing discharge typical in the Mega Ampere Spherical Tokamak (MAST) [Sykes et al., Nucl. Fusion 41, 1423 (2001)]. It is found that the ideal n=1 internal kink mode can be stabilized by toroidal rotation at values observed experimentally.
Residual gas analysis for long-pulse, advanced tokamak operation.
Klepper, C C; Hillis, D L; Bucalossi, J; Douai, D; Oddon, P; Vartanian, S; Colas, L; Manenc, L; Pégourié, B
2010-10-01
A shielded residual gas analyzer (RGA) system on Tore Supra can function during plasma operation and is set up to monitor the composition of the neutral gas in one of the pumping ducts of the toroidal pumped limited. This "diagnostic RGA" has been used in long-pulse (up to 6 min) discharges for continuous monitoring of up to 15 masses simultaneously. Comparison of the RGA-measured evolution of the H(2)/D(2) isotopic ratio in the exhaust gas to that measured by an energetic neutral particle analyzer in the plasma core provides a way to monitor the evolution of particle balance. RGA monitoring of corrective H(2) injection to maintain proper minority heating is providing a database for improved ion cyclotron resonance heating, potentially with RGA-base feedback control. In very long pulses (>4 min) absence of significant changes in the RGA-monitored, hydrocarbon particle pressures is an indication of proper operation of the actively cooled, carbon-based plasma facing components. Also H(2) could increase due to thermodesorption of overheated plasma facing components.
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.
Furth, H.P.
1984-10-01
The economic prospects of the tokamak are reviewed briefly and found to be favorable - if the size of ignited tokamak plasmas can be kept small and appropriate auxiliary systems can be developed. The main objectives of the Princeton Plasma Physics Laboratory tokamak program are: (1) exploration of the physics of high-temperature toroidal confinement, in TFTR; (2) maximization of the tokamak beta value, in PBX; (3) development of reactor-relevant rf techniques, in PLT.
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.
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. __________________________________________________
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.
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)
Zhou, Deng
2016-10-01
The dispersion relation of geodesic acoustic modes with a magnetic perturbation in the tokamak plasma with an equilibrium radial electric field was derived. The dispersion relation was analyzed for very low field strength. The mode frequency decreases with increasing field strength, which is different from the electrostatic geodesic acoustic mode. There exists an m = 1 magnetic component that is very low when the radial electric field is absent. The ratio between the m = 1 and m = 2 magnetic components increases with strength of the radial electric field for low Mach numbers.
Particle Confinement Properties of Lower Hybrid Current Drive Plasma on the HL-1 Tokamak
NASA Astrophysics Data System (ADS)
Duan, Xuru; Yuan, Chengjie; Qian, Shangjie; Ding, Xuantong; Yuan, Bin; Yang, Guang; Diao, Guangyue
1994-03-01
The particle confinement property of LHCD (lower hybrid current drive) plasma on the HL-1 tokamak is mainly affected by the line-averaged density of electrons (ne). With ne < 2.0 × 1013 cm-3, the particle confinement time (τp) is improved with the suppression of Hα(Dα) fluctuation at the edge, and tends to increase with the power PLH. The peak of τp appears near the critical density (1.0×1013 cm-3). These results are not influenced by the current drive directions.
de Baar, M.R.; Hogeweij, G.M.; Lopes Cardozo, N.J.; Oomens, A.A.; Schueller, F.C.
1997-06-01
In the Rijnhuizen Tokamak Project, plasmas with steady-state negative central shear (NCS) are made with off-axis electron cyclotron heating. Shifting the power deposition by 2mm results in a sharp transition of confinement. The good confinement branch features a transport barrier at the off-axis minimum of the safety factor (q) , where q{le}3, and two magnetohydrodynamic (MHD) instabilities, where one is localized at the off-axis minimum of q and one covers the entire NCS region. The low confinement branch has q{gt}3 everywhere, no transport barrier, and no MHD activity. {copyright} {ital 1997} {ital The American Physical Society}
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
NASA Astrophysics Data System (ADS)
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.
2013-09-01
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 q0<1, rotation increments in the countercurrent direction are observed. If the LH-driven current is sufficient to suppress sawteeth and increase q0 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.
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.
Shaping of the plasma column in a small aspect ratio tokamak
NASA Astrophysics Data System (ADS)
Herrera, Julio; Arroyo, Ismael; Chavez, Esteban; Segura, Miguel Angel
2016-10-01
This is a follow-up to the work presented in a precious 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.
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.
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.
Linear microstability analysis of a low-Z impurity doped tokamak plasma
NASA Astrophysics Data System (ADS)
Romanelli, M.; Szepesi, G.; Peeters, A. G.; Apicella, M. L.; Marinucci, M.; Mazzotta, C.; Mazzitelli, G.; Frigione, D.
2011-10-01
Improved electron and deuterium energy and particle confinement in the presence of low-Z impurities have been observed in many tokamaks under various experimental conditions. Peaked electron density profiles have been obtained in the Frascati Tokamak Upgrade (FTU) ohmic plasmas where a high concentration of lithium has been detected following the installation of a Liquid Lithium Limiter (LLL). This paper presents the results of a gyrokinetic study on the effects of lithium and other low-Z impurities on the linear stability of deuterium and electron temperature driven modes and their associated fluxes for plasma parameters such as those found in the core of LLL-FTU plasmas. Simulations show that a lithium concentration in excess of nLi/ne = 15%, as estimated in the initial phase of a reference FTU discharge, is found to have a strong stabilizing effect on the TEM and high-frequency ETG modes. A significant stabilization of the electron driven modes can still be observed when the lithium concentration is reduced to 3%. In the presence of a significant impurity concentration (nLi/ne = 3-15%) the long wavelength ITG modes drive an inward electron and deuterium flux and outward lithium flux. This process may lead eventually to an increased electron and deuterium density peaking and a reduced Zeff (lithium density below nLi/ne = 1%).
Inductive plasma current start-up by the outer vertical field coil in a spherical tokamak
NASA Astrophysics Data System (ADS)
Mitarai, Osamu
1999-12-01
Plasma current-start up induced by an outer vertical field coil is studied during the ignition access phase in a spherical tokamak reactor. We have illustrated the concept that the plasma current of ~50 MA could be induced by the outer vertical field coil in the proposed spherical tokamak with the help of the small central solenoid flux of +/-5 V s and the strong heating power less than 100 MW for the internal inductance of icons/Journals/Common/ell" ALT="ell" ALIGN="TOP"/>i~0.4-0.8 without the help of bootstrap current and non-inductive current drive power. The required condition to achieve this operation scenario is that the flux produced by the equilibrium vertical field is larger than the inductive flux. Current start-up operation is achieved by adding the small ohmic heating solenoid flux for the flux waveform adjustment because the flux from the outer vertical field coil cannot solely induce the desired plasma current waveform in the case of the preprogramming of the heating power.
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.
New fluctuation phenomena in the H-mode regime of PDX tokamak plasmas
Slusher, R.E.; Surko, C.M.; Valley, J.F.; Crowley, T.; Mazzucato, E.; McGuire, K.
1984-05-01
A new kind of quasi-coherent fluctuation is observed near the edge of plasmas in the PDX tokamak during H-mode operation. (The H-mode occurs in neutral beam heated divertor plasmas and is characterized by improved energy containment as well as large density and temperature gradients near the plasma edge.) These fluctuations are evidenced as VUV and density fluctuation bursts at well-defined frequencies (..delta omega../..omega.. less than or equal to 0.1) in the frequency range between 50 and 180 kHz. They affect the edge temperature-density product, and therefore they may be important for understanding the relationship between the large edge density and temperature gradients and the improved energy confinement.
NASA Astrophysics Data System (ADS)
Brunner, S.; Fivaz, M.; Tran, T. M.; Vaclavik, J.
1998-11-01
A solution to the full two-dimensional eigenvalue problem of electrostatic microinstabilities in a tokamak plasma is presented in the framework of gyrokinetic theory. The approach is the generalization of methods previously developed for a cylindrical system [S. Brunner and J. Vaclavik, Phys. Plasmas 5, 365 (1998)]. By solving the spectral problem in a special Fourier space adapted to the curved geometry, orbit width as well as Larmor radius can be kept to all orders. For a first numerical implementation, a large aspect ratio plasma with circular concentric magnetic surfaces is considered. A root finding algorithm for identifying the eigenfrequencies, based on a higher order Nyquist method, enables straightforward implementation on a parallel computer. Illustrative results for ion temperature gradient-related instabilities are presented. These include scaling studies of the radial width, and toroidicity and magnetic shear scans, as well as the effects of nonadiabatic trapped electron dynamics.
Physics of High-Power ECH Plasmas in T-10 Tokamak
Kislov, D. A.
2006-01-15
Physics of plasma confinement and stability under the conditions of electron cyclotron heating (ECH) is under investigation in T-10 tokamak. High-density plasmas with energy confinement time that exceeds the H-mode scaling predictions have been obtained both with gas puffing and with deuterium pellet injection. Transient internal transport barrier formation has been observed with ECH during the current ramp-up and after off-central ECH switch off. A systematic study of plasma turbulence in a wide range of operating regimes has been performed and a possible link between transport and turbulence properties is under consideration. The value of critical for neoclassical tearing mode onset beta was found to be dependent on q(r) profile. Physical mechanism of sawtooth control by highly localized ECH is analyzed.
A study of quasi-mode parametric excitations in lower-hybrid heating of tokamak plasmas
NASA Astrophysics Data System (ADS)
Villalon, E.; Bers, A.
1980-03-01
A detailed linear and non-linear analysis of quasi-mode parametric excitations, relevant to experiments in supplementary heating of tokamak plasmas, is presented. The linear analysis includes the full ion-cyclotron harmonic quasi-mode spectrum, while the nonlinear one, considering depletion of the pump electric field, is applied to the recent Alcator A heating experiment. The quasi-mode excitations are studied independently for the plasma edge and the main bulk of the plasma, and for the two typical regimes in overall density. It is concluded that the excited spectrum has a frequency close to the initial pump frequency, while the wave-number spectrum may be different from the initial linear spectrum.
Hassanein, A.; Konkashbaev, I.
1999-10-25
Loss of plasma confinement causes surface and structural damage to plasma-facing materials (PFMs) and remains a major obstacle for tokamak reactors. The deposited plasma energy results in surface erosion and structural failure. The surface erosion consists of vaporization, spallation, and liquid splatter of metallic materials, while the structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. Comprehensive models (contained in the HEIGHTS computer simulation package) are being used self-consistently to evaluate material damage. Splashing mechanisms occur as a result of volume bubble boiling and liquid hydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials. The effect of macroscopic erosion on total mass losses and lifetime is evaluated. The macroscopic erosion products may further protect PFMs from severe erosion (via the droplet-shielding effect) in a manner similar to that of the vapor shielding concept.
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.
Estimation of the current driven by residual loop voltage in LHCD plasma on EAST Tokamak
NASA Astrophysics Data System (ADS)
Zhang, X. M.; Yu, L. M.; Wan, B. N.; Xue, E. B.; Fang, Y.; Shi, K. Y.; EAST Team
2016-02-01
The lower hybrid wave current drive (LHCD) is one of the efficient methods of driving the non-inductive current required for Tokamak operating in steady-state. Residual loop voltage exists in Tokamak when the non-inductive current is not fully driven. Residual loop voltage also accelerates the fast electrons generated by the lower hybrid wave (LHW), which can drive extra current and combine with the current driven by the LHW. It is generally difficult to separate these two different components of driven current in the experiment. In this paper, the currents driven by LHCD and residual loop voltage are separated directly by solving the Fokker-Plank equation numerically. The fraction of the current driven by residual loop voltage compared to the current driven by LHW is evaluated on the experimental advanced superconducting tokamak (EAST). The current driven by residual loop voltage is several percent of the currents driven by the LHCD when the residual loop voltage is small, but it increases with the residual loop voltage up to 25% when the residual loop voltage is about 2 V. The hot electrical conductivity is deduced from the net current driven by the residual loop voltage. Its distribution profile is related to the fast electron distribution driven by LHW.
Advanced Computation in Plasma Physics
NASA Astrophysics Data System (ADS)
Tang, William
2001-10-01
Scientific simulation in tandem with theory and experiment is an essential tool for understanding complex plasma behavior. This talk will review recent progress and future directions for advanced simulations in magnetically-confined plasmas with illustrative examples chosen from areas such as microturbulence, magnetohydrodynamics, magnetic reconnection, and others. Significant recent progress has been made in both particle and fluid simulations of fine-scale turbulence and large-scale dynamics, giving increasingly good agreement between experimental observations and computational modeling. This was made possible by innovative advances in analytic and computational methods for developing reduced descriptions of physics phenomena spanning widely disparate temporal and spatial scales together with access to powerful new computational resources. In particular, the fusion energy science community has made excellent progress in developing advanced codes for which computer run-time and problem size scale well with the number of processors on massively parallel machines (MPP's). A good example is the effective usage of the full power of multi-teraflop MPP's to produce 3-dimensional, general geometry, nonlinear particle simulations which have accelerated progress in understanding the nature of turbulence self-regulation by zonal flows. It should be emphasized that these calculations, which typically utilized billions of particles for tens of thousands time-steps, would not have been possible without access to powerful present generation MPP computers and the associated diagnostic and visualization capabilities. In general, results from advanced simulations provide great encouragement for being able to include increasingly realistic dynamics to enable deeper physics insights into plasmas in both natural and laboratory environments. The associated scientific excitement should serve to stimulate improved cross-cutting collaborations with other fields and also to help attract
Advanced computations in plasma physics
NASA Astrophysics Data System (ADS)
Tang, W. M.
2002-05-01
Scientific simulation in tandem with theory and experiment is an essential tool for understanding complex plasma behavior. In this paper we review recent progress and future directions for advanced simulations in magnetically confined plasmas with illustrative examples chosen from magnetic confinement research areas such as microturbulence, magnetohydrodynamics, magnetic reconnection, and others. Significant recent progress has been made in both particle and fluid simulations of fine-scale turbulence and large-scale dynamics, giving increasingly good agreement between experimental observations and computational modeling. This was made possible by innovative advances in analytic and computational methods for developing reduced descriptions of physics phenomena spanning widely disparate temporal and spatial scales together with access to powerful new computational resources. In particular, the fusion energy science community has made excellent progress in developing advanced codes for which computer run-time and problem size scale well with the number of processors on massively parallel machines (MPP's). A good example is the effective usage of the full power of multi-teraflop (multi-trillion floating point computations per second) MPP's to produce three-dimensional, general geometry, nonlinear particle simulations which have accelerated progress in understanding the nature of turbulence self-regulation by zonal flows. It should be emphasized that these calculations, which typically utilized billions of particles for thousands of time-steps, would not have been possible without access to powerful present generation MPP computers and the associated diagnostic and visualization capabilities. In general, results from advanced simulations provide great encouragement for being able to include increasingly realistic dynamics to enable deeper physics insights into plasmas in both natural and laboratory environments. The associated scientific excitement should serve to
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.
Comparison of dust transport modelling codes in a tokamak plasma
NASA Astrophysics Data System (ADS)
Uccello, Andrea; Gervasini, Gabriele; Ghezzi, Francesco; Lazzaro, Enzo; Bacharis, Minas; Flanagan, Joanne; Matthews, Guy; Järvinen, Aaro; Sertoli, Marco
2016-10-01
Since the installation on the Joint European Torus of the ITER-like Wall (ILW), intense radiation spikes have been observed, especially in the discharges following a disruption, and have been associated with possible sudden injection of tungsten (W) impurities consequent to full ablation of W dust particles. The problem of dust production, mobilization, and interaction both with the plasma and the vessel tiles is therefore of great concern and requires the setting up of dedicated and validated numerical modeling tools. Among these, a useful role is played by the dust trajectory calculators, which can present in a relatively clear way the qualitative and quantitative description of the mobilization and fate of selected bunches of dust grains.
Fast tokamak plasma flux and electron density reconstruction technique
Chiang, K.L.; Hallock, G.A.; Wootton, A.J.; Wang, L.
1997-01-01
Density profiles in TEXT-U are obtained using a vertical viewing far-infrared (FIR) interferometer. To obtain the local (inverted) density, we have developed a simple analytic model of the plasma equilibrium configuration which is faster than EFIT (a flux surface reconstruction program) and can be easily computed between discharges. This analytic solution of the Grad{endash}Shafranov equation is valid as long as the pressure p is a function of poloidal flux {psi}, i.e., p=p({psi}). The procedure incorporates both magnetic and FIR density data to solve the Grad{endash}Shafranov equation, and provides a density profile which is self-consistent with the reconstructed equilibrium flux surfaces. Examples are presented. {copyright} {ital 1997 American Institute of Physics.}
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.
Myra, J.R.; Catto, P.J. ); Wootton, A.J.; Bengtson, R.D. ); Wang, P.W. )
1992-07-01
Recent experimental and theoretical results are combined to explore the energy dependence of runaway and suprathermal electron diffusion. It is shown that a quasilinear theory containing both magnetic and electrostatic {ital E}{times}{ital B}-driven radial transport is consistent with the available experimental data, and supports the notion that magnetic turbulence dominates runaway diffusion but electrostatic turbulence dominates thermal diffusion in the edge plasma of the Texas Experimental Tokamak (TEXT) (Nucl. Technol./Fusion {bold 1}, 479 (1981)). These observations further expand the possibilities for employing runaway or suprathermal electrons as a diagnostic of the underlying transport mechanisms in a tokamak plasma.
NASA Astrophysics Data System (ADS)
Liu, D. M.; Li, J.; Wan, B. N.; Lu, Z.; Wang, L. S.; Jiang, L.; Lu, C. H.; Huang, J.
2016-11-01
As one of the core subsystems of the Experimental Advanced Superconducting Tokamak (EAST), the poloidal field power system supplies energy to EAST's superconducting coils. To measure the converter current in the poloidal field power system, a current measurement system has been designed. The proposed measurement system is composed of a Rogowski coil and a newly designed integrator. The results of the resistor-inductor-capacitor discharge test and the converter equal current test show that the current measurement system provides good reliability and stability, and the maximum error of the proposed system is less than 1%.
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.
Liu, D M; Li, J; Wan, B N; Lu, Z; Wang, L S; Jiang, L; Lu, C H; Huang, J
2016-11-01
As one of the core subsystems of the Experimental Advanced Superconducting Tokamak (EAST), the poloidal field power system supplies energy to EAST's superconducting coils. To measure the converter current in the poloidal field power system, a current measurement system has been designed. The proposed measurement system is composed of a Rogowski coil and a newly designed integrator. The results of the resistor-inductor-capacitor discharge test and the converter equal current test show that the current measurement system provides good reliability and stability, and the maximum error of the proposed system is less than 1%.
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
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
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.
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).
TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry
X. Q. Xu; Bodi, K.; Cohen, R. H.; ...
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
Modeling of runaway electron damage for the design of tokamak plasma facing components
Niemer, K.A.; Gilligan, J.G. . Dept. of Nuclear Engineering); Croessmann, C.D. ); Bolt, H.H. . NET Design Team)
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/m{sup 2} 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. 9 refs., 7 figs.
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.
Henriques, R B; Malaquias, A; Nedzelskiy, I S; Silva, C; Coelho, R; Figueiredo, H; Fernandes, H
2014-11-01
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 neσ1,2(Te) 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 neσ1,2(Te) fluctuations.
Initial confinement studies of ohmically heated plasmas in the Tokamak Fusion Test Reactor
Efthimion, P.C.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.L.; Coonrod, J.; Davis, S.; Dylla, H.F.; Fonck, R.; Furth, H.P.
1984-06-01
Initial operation of the Tokamak Fusion Test Reactor (TFTR) has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1 to 0.2 s for a line-average density range (anti n/sub e/) of 1 to 2.5 x 10/sup 19/ m/sup -3/ with electron temperatures of T/sub e/(o) approx. 1.2 to 2.2 keV, ion temperatures of T/sub i/(o) approx. 0.9 to 1.5 keV, and Z/sub eff/ approx. 3. A comparison of PLT, PDX, and TFTR plasma confinement supports a dimension-cubed scaling law.
Observation of ion-cyclotron-frequency mode-conversion flow drive in tokamak plasmas.
Lin, Y; Rice, J E; Wukitch, S J; Greenwald, M J; Hubbard, A E; Ince-Cushman, A; Lin, L; Porkolab, M; Reinke, M L; Tsujii, N
2008-12-05
Strong toroidal flow (Vphi) and poloidal flow (Vtheta) have been observed in D-3He plasmas with ion cyclotron range of frequencies (ICRF) mode-conversion (MC) heating on the Alcator C-Mod tokamak. The toroidal flow scales with the rf power Prf (up to 30 km/s per MW), and is significantly larger than that in ICRF minority heated plasmas at the same rf power or stored energy. The central Vphi responds to Prf faster than the outer regions, and the Vphi(r) profile is broadly peaked for r/a < or =0.5. Localized (0.3 < or = r/a < or =0.5) Vtheta appears when Prf > or =1.5 MW and increases with power (up to 0.7 km/s per MW). The experimental evidence together with numerical wave modeling suggests a local flow drive source due to the interaction between the MC ion cyclotron wave and 3He ions.
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.
Observation of anomalous momentum transport in tokamak plasmas with no momentum input.
Lee, W D; Rice, J E; Marmar, E S; Greenwald, M J; Hutchinson, I H; Snipes, J A
2003-11-14
Anomalous momentum transport has been observed in Alcator C-Mod tokamak plasmas through analysis of the time evolution of core impurity toroidal rotation velocity profiles. Following the L-mode to EDA (enhanced D(alpha)) H-mode transition, the ensuing cocurrent toroidal rotation velocity, which is generated in the absence of any external momentum source, is observed to propagate in from the edge plasma to the core. The steady state toroidal rotation velocity profiles are relatively flat and the momentum transport can be simulated with a simple diffusion model. Velocity profiles during edge localized mode free (ELM-free) H-modes are centrally peaked, which suggests the addition of inward momentum convection. In all operating regimes the observed momentum diffusivities are much larger than the neoclassical values.
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.
Measurement of the edge plasma rotation on J-TEXT tokamak
NASA Astrophysics Data System (ADS)
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.
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)
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.
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.
Low-frequency fluctuations in scrape-off layer of tokamak plasma with limiter biasing
NASA Astrophysics Data System (ADS)
Bisai, N.; Singh, Rameswar; Singh, R.
2011-12-01
The effects of limiter biasing on the equilibrium density and potential profiles of the scrape-off layer (SOL) in tokamak plasma are investigated by including ionization and cross-field mobility. It is shown that a broadening of SOL can take place by the inclusion of ionization for low negative biasing. Various microinstabilities relevant for SOL plasmas have been studied. Generalized low-frequency dispersion relation is derived. It is shown that limiter biasing significantly modifies the SOL fluctuations. It is also shown that growth rate of conductive wall instability is smaller for negative biasing than positive biasing case. New mode, the modified Simon-Hoh, and ion temperature gradient instabilities are found to contribute significantly to the growth of curvature- and electron temperature-gradient-driven conductive wall instabilities.
Ion-Bernstein-wave heating in the JIPPT-II-U tokamak plasma
NASA Astrophysics Data System (ADS)
Ono, M.; Watari, T.; Ando, R.; Fujita, J.; Hirokura, Y.; Ida, K.; Kako, E.; Kawahata, K.; Kawasumi, Y.; Matsuoka, K.; Nishizawa, A.; Noda, N.; Ogawa, I.; Ohkubo, K.; Okamoto, M.; Soto, K.; Tanahashi, S.; Taniguchi, Y.; Tetsuka, T.; Toi, K.; Yamazaki, K.
1985-05-01
Ion-Bernstein-wave heating is investigated in the JIPPT-II-U tokamak plasma, n¯e~=1.5×1013 cm-3, Te0~=700 eV, and Ti0~=300 eV for Prf<~100 kW. When the (3/2)ΩH layer is placed near the plasma minor axis, the bulk hydrogen-ion temperature shows a significant rise, ΔTi⊥<=700 eV and ΔTipara<=300 eV. The ion heating dependence on the magnetic field and rf power suggests a presence of direct bulk hydrogen heating mechanism at ω~=(3/2)ΩH.
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.
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.
Compressional Alfvén eigenmodes in rotating spherical tokamak plasmas
NASA Astrophysics Data System (ADS)
Smith, H. M.; Fredrickson, E. D.
2017-03-01
Spherical tokamaks often have a considerable toroidal plasma rotation of several tens of kHz. Compressional Alfvén eigenmodes in such devices therefore experience a frequency shift, which if the plasma were rotating as a rigid body, would be a simple Doppler shift. However, since the rotation frequency depends on minor radius, the eigenmodes are affected in a more complicated way. The eigenmode solver CAE3B (Smith et al 2009 Plasma Phys. Control. Fusion 51 075001) has been extended to account for toroidal plasma rotation. The results show that the eigenfrequency shift due to rotation can be approximated by a rigid body rotation with a frequency computed from a spatial average of the real rotation profile weighted with the eigenmode amplitude. To investigate the effect of extending the computational domain to the vessel wall, a simplified eigenmode equation, yet retaining plasma rotation, is solved by a modified version of the CAE code used in Fredrickson et al (2013 Phys. Plasmas 20 042112). In summary, both solving the full eigenmode equation, as in the CAE3B code, and placing the boundary at the vessel wall, as in the CAE code, significantly influences the calculated eigenfrequencies.
Two-dimensional MHD simulations of tokamak plasmas with poloidal flow
NASA Astrophysics Data System (ADS)
Hu, Bo; Betti, R.
2006-10-01
It has been shown [1] that, according to the ideal MHD equilibrium theory, poloidal flow in a tokamak can give rise to a pedestal structure with the pressure, density and velocity developing sharp discontinuities in their radial profiles. Such a pedestal arises when the poloidal velocity exceeds the poloidal sound speed. Since the poloidal sound speed vanishes at the separatrix, it is conceivable that evena rather slow poloidal flow can become transonic near the plasma edge, thus inducing a pedestal in the hydrodynamic profiles. While equilibrium calculations [1-4] of such a pedestal are well established, only a few two-dimensional time-dependent simulations have been carried out [5]. Here, we show the preliminary results from a two dimensional MHD code that simulates the formation of the pedestal starting from a poloidal velocity profile that becomes supersonic at the plasma edge. This work was supported by US-DOE under Contract DE-FG02-93ER54215. [1] Betti and Freidberg, Phys. Plasmas 7, 2439 (2000). [2] Guazzotto, Betti, Manickam and Kaye, Phys. Plasmas 11, 604 (2004). [3] Guazzotto and Betti, Phys. Plasmas 12, 056107 (2005). [4] Thyagaraja and McClements, Phys. Plasmas 13, 062502 (2006). [5] Gardiner, Betti and Guazzotto, Bull. Am. Phys. Soc. 46, No. 8, 166 (2001).
Kim, Kimin; Ahn, J. -W.; Scotti, F.; ...
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
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
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.
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.
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.
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)
Yan, Xingting; Zhu, Ping; Sun, Youwen
2016-10-01
The characteristic profile and magnitude are predicted in theory for the neoclassical toroidal viscosity (NTV) torque induced by the plasma response to the resonant magnetic perturbation (RMP) in a tokamak with an edge pedestal, using the newly developed module coupling the NIMROD and the NTVTOK codes. For a low β equilibrium, the NTV torque is mainly induced by the dominant toroidal mode of plasma response. The NTV torque profile is radially localized and peaked, which is determined by profiles of both the equilibrium temperature and the plasma response fields. In general, the peak of NTV torque profile is found to trace the pedestal location. The magnitude of NTV torque is extremely sensitive to the β of pedestal top; for a given plasma response, the peak value of NTV torque can increase by three orders of magnitude, when the pedestal β increases by only one order of magnitude. This suggests a more significant role of NTV torque in higher plasma β regimes. Supported by the National Magnetic Confinement Fusion Program of China under Grant Nos. 2014GB124002 and 2015GB101004, and the 100 Talent Program of the Chinese Academy of Sciences.
Investigation on the effect of pressure on turbulent transports of the IR-T1 Tokamak plasma
NASA Astrophysics Data System (ADS)
Alipour, Ramin; Meshkani, Sakineh; Elahi, Ahmad Salar; Ghoranneviss, Mahmood
2017-03-01
Investigation on the effects of limiter biasing and applying external electric fields (positive and negative voltage of 200 V) under various pressures of 1.9, 2.3 and 2.7 Torr on the IR-T1 Tokamak plasma parameters such as plasma current, loop voltage, radial electric field, poloidal electric field and Reynolds stress and turbulent transports, was done in this work. The results show that the ramp time of the plasma current decreased and the change rate of loop voltage raised by increasing plasma pressure. The power of radial electric field is more than the power of poloidal electric field. According to the applied positive voltage to the plasma, it is observed that the Reynolds stress of IR-T1 Tokamak plasma has experienced a more significant reduction at the pressure of 2.3 Torr than other pressures. The results Also show that the applying positive voltage of 200 V to the IR-T1 Tokamak plasma at the pressure of 2.3 Torr (compared to other pressures) has more effect on increases and decreases the poloidal and radial turbulent transport, respectively.
Circular limiter H-mode plasmas in the Tokamak Fusion Test Reactor (TFTR)
Bush, C.E.
1990-01-01
Circular limiter H-modes with centrally peaked density profiles have been obtained in TFTR using a highly conditioned graphite limiter. The transition to these centrally peaked H-modes takes place from the supershot to the H-mode rather than the usual L- to H-mode transition observed in other tokamaks. Bidirectional beam heating is required and the threshold power needed to induce the transition increases linearly with plasma current. Density peaking factors, n{sub e}(0)/{l angle}n{sub e}{r angle}, greater than 2.3 are obtained and, at the same time, the H-mode characteristics are similar to those of limiter H-modes on other tokamaks and the global confinement, {tau}{sub E}, can be >2.5 times L-mode scaling. Microwave scattering data from the edge plasma shows broad spectra at k = 5.5 cm{sup {minus}1} which begin at the drop in D{sub {alpha}} radiation and are strongly shifted in the electron diamagnetic drift direction. This implies a poloidal rotation, which begins at the transition to the H-mode, of {approximately}10{sup 4} m/sec. During an edge localized mode instability (ELM), these apparent rotations cease and Mirnov fluctuations in the 50--500 kHz range increase in intensity. Electron cyclotron emission data shows the origin of the ELMs and probably the transition layer to be located a few centimeters inside the plasma surface. A short review of requirements for controlled thermonuclear reactions is given in the introduction. 16 refs., 7 figs.
NASA Astrophysics Data System (ADS)
Ming, Yue; Zhou, Deng
2017-01-01
The effect of the poloidal equilibrium flow and flow shear on the tearing mode instabilities for tokamak plasmas is investigated. The vorticity equation is derived and approximately solved for large poloidal mode numbers (m). Asymptotic matching of the inner solution to the outer solution can approximately give the classical tearing mode stability index Δ' . For typical plasma parameters with positive flow shear, we notice that the poloidal mean flows have a beneficial effect on the classical tearing mode and vice versa. To study the modes with arbitrary poloidal mode numbers, we numerically solve the vorticity equation for delta prime ( Δ' ) for typical plasma parameters with positive flow shear at the rational surface and the resulting Δ' with large m also decreases with increasing poloidal flow velocity, consistent with the approximate analytical large m results. Our numerical calculations indicate that the poloidal mean flow with positive flow shear has beneficial influence on the stabilization of classical tearing modes in tokamak plasmas.
Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.
1981-10-01
Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.
NASA Astrophysics Data System (ADS)
Zhang, J. Z.; Zhu, Y. B.; Zhao, J. L.; Wan, B. N.; Li, J. G.; Heidbrink, W. W.
2016-11-01
Full function integrated, compact solid state neutral particle analyzers (ssNPA) based on absolute extreme ultraviolet silicon photodiode have been successfully implemented on the experimental advanced superconducting tokamak to measure energetic particle. The ssNPA system has been operated in advanced current mode with fast temporal and spatial resolution capabilities, with both active and passive charge exchange measurements. It is found that the ssNPA flux signals are increased substantially with neutral beam injection (NBI). The horizontal active array responds to modulated NBI beam promptly, while weaker change is presented on passive array. Compared to near-perpendicular beam, near-tangential beam brings more passive ssNPA flux and a broader profile, while no clear difference is observed on active ssNPA flux and its profile. Significantly enhanced intensities on some ssNPA channels have been observed during ion cyclotron resonant heating.
Zhang, J Z; Zhu, Y B; Zhao, J L; Wan, B N; Li, J G; Heidbrink, W W
2016-11-01
Full function integrated, compact solid state neutral particle analyzers (ssNPA) based on absolute extreme ultraviolet silicon photodiode have been successfully implemented on the experimental advanced superconducting tokamak to measure energetic particle. The ssNPA system has been operated in advanced current mode with fast temporal and spatial resolution capabilities, with both active and passive charge exchange measurements. It is found that the ssNPA flux signals are increased substantially with neutral beam injection (NBI). The horizontal active array responds to modulated NBI beam promptly, while weaker change is presented on passive array. Compared to near-perpendicular beam, near-tangential beam brings more passive ssNPA flux and a broader profile, while no clear difference is observed on active ssNPA flux and its profile. Significantly enhanced intensities on some ssNPA channels have been observed during ion cyclotron resonant heating.
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.
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.
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.
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.
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.
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.
Design of divertor plate and measurements of double-null open divertor plasma in the JFT-2M tokamak
NASA Astrophysics Data System (ADS)
Yanagisawa, Ichiro; Shoji, Teruaki; Mori, Masahiro; Odajima, Kazuo; Ohtsuka, Hideo; Suzuki, Norio; Hasegawa, Mitsuru; Ohta, Kanji; Sugihara, Masayoshi; Uesugi, Yoshihiko
1987-10-01
The Design of the divertor plate, the results of the computational simulation and the experimental results on the compact diverter of the JFT-2 tokamak are described. Graphite divertor plates have showed a good performance as divertor target materials through divertor discharges. The H-mode plasma and low temperature, high density divertor plasma are obtained. From computational results, this is in the intermediate region between low and high recycling region.
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.
The influence of electrode biasing on plasma confinement in the J-TEXT tokamak
NASA Astrophysics Data System (ADS)
Sun, Yue; Chen, Z. P.; Zhu, T. Z.; Yu, Q.; Zhuang, G.; Nan, J. Y.; Ke, X.; Liu, H.; the J-TEXT Team
2014-01-01
The influence of both positive and negative bias on global and plasma-edge parameters has been comparatively studied with a newly designed electrode biasing system in the J-TEXT tokamak. Compared to the 0 V bias case, the global particle confinement of plasma is enhanced under bias with both polarities, with the increments of the central line-averaged density and the soft x-ray emission, as well as the reduction of the edge Hα radiation level. The suppression of plasma-edge fluctuations and turbulent particle transport are obviously observed under bias, in different degrees with different polarities. The potential fluctuation amplitude is observed to be increased at the vicinity of the limiter under positive bias, with the existence of a peaked low-frequency mode characterized as high coherence and near-zero cross-phase poloidally in the edge region, which is not found in the negative bias case. The poloidal correlation length of turbulence is greatly enhanced under bias with both polarities; it shows a positive correlation with the amplitude of the poloidal phase velocity, which is mainly driven by the local Jr × B torque at the plasma edge under bias. The characteristic parameters of intermittent events (i.e. blobs), including amplitude, radial velocity, related particle flux and radial size, decreased dramatically under bias in the edge region.
Two-Dimensional MHD Simulations of Tokamak Plasmas with Poloidal Flow
NASA Astrophysics Data System (ADS)
Guazzotto, L.; Betti, R.
2002-11-01
A two- dimensional MHD code has been developed to simulate the temporal evolution of Tokamak plasmas with an imposed poloidal flow. The code is fully compressible and can resolve the shock structures arising when the poloidal velocity is of the order of the poloidal sound speed (V_θ ˜ Cs B_θ/B) near the plasma edge, where the plasma is cold and the sound speed is low. The poloidal flow is assigned as an initial condition with a velocity profile ranging from subsonic to supersonic near the edge. It is found that a continuous band of shocks is formed near the edge. Such shocks travel poloidally, leaving behind a pedestal structure similar to the one predicted in Ref. 1 [R. Betti and J. P. Freidberg, Phys. Plasmas 7, 2439 (2000)]. Here, the pedestal is defined as a sharp discontinuity in the pressure, temperature, and density profiles. The pedestal height is modulated in the poloidal angle; it is maximum on the outboard side (θ = 0) and minimum on the inboard (θ = π). Furthermore, both poloidal and toroidal flows develop a shear layer at the location of the pedestal. The large velocity shear (both poloidal and toroidal) occurring in the pedestal region is likely to suppress turbulent eddies and reduce anomalous transport. This work was supported by the U.S. Department of Energy Office of Inertial Confinement Fusion under Cooperative Agreement No. DE-FC03-92SF19460.
Laser Induced Plasma Spectroscopy to Diagnose Impurities on a Tokamak Divertor
NASA Astrophysics Data System (ADS)
Kim, Minju; Cho, Min Sang; Cho, Byoung-Ick
2015-11-01
In order to monitor dust and impurity deposition on the plasma facing components (PFCs) of a fusion device, the Laser Induced Plasma Spectroscopy (LIPS) is considered. It is a powerful spectroscopic technique to measure emission lines from the excited atoms by means of the high power laser pulse, and could be applied to diagnose dust and impurities deposition on the PFCs. We have measured LIPS spectra for the inner-divertor tile from 2011 KSTAR campaign. Characteristic emission lines for several key elements, such as iron, chrome are identified. Using those lines, plasma conditions for various laser parameters and their temporal evolution are characterized. It will be also presented that the depth profiling for the deposited elements on a surface of graphite tile. This work is supported by the NRF (No. 2013M1A7A1A02043864), National Research Foundation of Korea (No. 2013R1A1A1007084) and the TBP research project of GIST. Laser Induced Plasma Spectroscopy to Diagnose Impurities on a Tokamak Divertor.
Gourdain, P-A; Peebles, W A
2008-10-01
Reflectometry has successfully demonstrated measurements of many important parameters in high temperature tokamak fusion plasmas. However, implementing such capabilities in a high-field, large plasma, such as ITER, will be a significant challenge. In ITER, the ratio of plasma size (meters) to the required reflectometry source wavelength (millimeters) is significantly larger than in existing fusion experiments. This suggests that the flow of the launched reflectometer millimeter-wave power can be realistically analyzed using three-dimensional ray tracing techniques. The analytical and numerical studies presented will highlight the fact that the group velocity (or power flow) of the launched microwaves is dependent on the direction of wave propagation relative to the internal magnetic field. It is shown that this dependence strongly modifies power flow near the cutoff layer in a manner that embeds the local magnetic field direction in the "footprint" of the power returned toward the launch antenna. It will be shown that this can potentially be utilized to locally determine the magnetic field pitch angle at the cutoff location. The resultant beam drift and distortion due to magnetic field and relativistic effects also have significant consequences on the design of reflectometry systems for large, high-field fusion experiments. These effects are discussed in the context of the upcoming ITER burning plasma experiment.
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.
Prospects for Tokamak Fusion Reactors
Sheffield, J.; Galambos, J.
1995-04-01
This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.
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.
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.
NASA Astrophysics Data System (ADS)
Fitzpatrick, Richard
2016-12-01
The simple analysis of Rutherford [Phys. Fluids 16, 1903 (1973)] is generalized in order to incorporate radial magnetic island asymmetry into the nonlinear theory of tearing mode stability in a low-β, large aspect-ratio, quasi-cylindrical, tokamak plasma. The calculation is restricted to cases in which the radial shifts of the island X- and O-points are (almost) equal and opposite. For the sake of simplicity, the calculation concentrates on a particular (but fairly general) class of radially asymmetric island magnetic flux-surfaces that can all be mapped to the same symmetric flux-surfaces by means of a suitable coordinate transform. The combination of island asymmetry (in which the radial shifts of the X- and O-points are almost equal and opposite) and temperature-induced changes in the inductive current profile in the immediate vicinity of the island is found to have no effect on tearing mode stability.
Neoclassical transport coefficients for finite-aspect-ratio and bean-shaped tokamak plasmas
NASA Astrophysics Data System (ADS)
Crume, E. C., Jr.; Beasley, C. O., Jr.; Hirshman, S. P.; van Rij, W. I.
1987-04-01
Numerically calculated tokamak equilibria are used to compute banana-plateau transport coefficients for finite-aspect-ratio, finite-beta plasmas. Calculations are presented for the Spherical Torus Experiment (STX) (NTIS Document No. DE 86004663) and the Princeton Beta Experiment (PBX) (NTIS Document No. DE 86011173). In STX, the poloidal variation of B≡‖B‖ over a magnetic surface tends to be reduced in regions of large major radius R. The reduction of radial transport caused by this quasiomnigeneous condition is offset by increased drifts and trapping probabilities for smaller R. Thus the modulation Δ=(Bmax-Bmin)/(Bmax+Bmin) on a magnetic surface becomes the critical parameter determining neoclassical transport. In PBX, the bean-shaped topology of the magnetic surfaces leads to the presence of multiple magnetic wells. Numerical calculations confirm that analytic calculations of neoclassical transport based on the total fraction of circulating particles are valid even when geometrically distinct classes of trapped particles are present.
Alfvén acoustic channel for ion energy in high-beta tokamak plasmas.
Bierwage, Andreas; Aiba, Nobuyuki; Shinohara, Kouji
2015-01-09
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.
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.
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.
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.
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
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.
NASA Astrophysics Data System (ADS)
Salar Elahi, A.; Ghoranneviss, M.
In this contribution we have presented a new perspective for determination of tokamak plasma column shift based on remote multipole moments technique. First, we presented analytical details for using this technique. Then principle of different models based on this technique for design and fabrication of a six coils will be presented: four modified Rogowski coil (two Cosine coils and two Sine coils) and two Saddle coils (Saddle Sine coil (SSC) and Saddle Cosine coil (SCC)). Also, in order to comparison of results the flux loops technique is used. Because of continuous measurements of magnetic field distribution around the tokamak plasma using multipole coils, this technique give us more reliable information about the plasma current displacement.
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.
Enhanced confinement regimes and control technology in the DIII-D tokamak
Lohr, J.; Burrell, K.H.; Coda, S.
1993-07-01
Advanced tokamak performance has been demonstrated in the DIII-D tokamak in a series of experiments which brought together developments in technology and improved understanding of the physical principles underlying tokamak operation. The achievement of greatly improved confinement coupled with development of new systems for real time plasma control have permitted investigation of the heretofore hidden or poorly controlled variables which together determine global confinement. These experiments, which included work in transport and control of the plasma boundary, point toward development of operationally and economically attractive reactors based on the tokamak. Some of these experiments are described.
NASA Astrophysics Data System (ADS)
Shaing, K. C.; Lee, H.; Seol, J.; Aydemir, A. Y.
2015-08-01
Theory for neoclassical toroidal plasma viscosity in the low collisionality regime is extended to the vicinity of the magnetic axis in tokamaks with broken symmetry. The toroidal viscosity is induced by particles drifting off the perturbed magnetic surface under the influence of the symmetry breaking magnetic field. In the region away from the magnetic axis, the drift orbit dynamics is governed by the bounce averaged drift kinetic equation in the low collisionality regimes. In the vicinity of the magnetic axis, it is the drift kinetic equation, averaged over the trapped particle orbits, i.e., potato orbits, that governs the drift dynamics. The orbit averaged drift kinetic equation is derived when collision frequency is low enough for trapped particles to complete their potato trajectories. The resultant equation is solved in the 1 /ν regime to obtain transport fluxes and, thus, toroidal plasma viscosity through flux-force relation. Here, ν is the collision frequency. The viscosity does not vanish on the magnetic axis, and has the same scalings as that in the region away from magnetic axis, except that the fraction of bananas is replaced by the fraction of potatoes. It also has a weak radial dependence. Modeling of plasma flow velocity V for the case where the magnetic surfaces are broken is also discussed.
Soft-computing approach to plasma evolution tracking in tokamak reactors
NASA Astrophysics Data System (ADS)
Morabito, Francesco C.
1997-10-01
Qualitative information about the structure of a mapping can surely be of help in learning a mapping by a collection of input-output pairs. However, there are conditions in which time and some other constraints make guessing the only plausible means for interpreting data. In this paper, the problem of the plasma boundary reconstruction in 'Tokamak' nuclear fusion rectors is assessed. The problem is formulated as an inverse 'identification' problem and the mapping is derived by a properly generated database of simulated experiments. Real data coming from experiments are also available to validate both numerically generated data and extracted model. The identification problem is solved for two different databases by using neural networks and more conventional models. The introduction of techniques derived from soft computing is shown to improve the performance in various respects. Dynamic identification systems appear to be rather demanding also for such systems, for the need of rapidly interpreting real time data for discharge control. Soft computing approaches may yet yield some low cost ways to take decisions during plasma evolution. The approximate analysis of experimental data could also improve the knowledge on the particular problem allowing an evolution of the knowledge base. Experimental data related to ASDEX-Upgrade machine are presented in this work and preliminary processed. Soft computing techniques also allow to simply get ideas about two other interesting problems in plasma engineering, namely, the fault tolerance and the minimization of the number of sensors.
Global, Gyrokinetic Eigenvalue Calculations of ITG-Related Microinstabilities in Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Brunner, S.; Fivaz, M.; Tran, T. M.; Vaclavik, J.
1998-11-01
Methods previously developed for a cylindrical system(S.Brunner and J.Vaclavik, Phys.Plasmas 5), 365 (1998) have been generalized to a tokamak plasma for solving the full 2-dimensional eigenvalue problem of electrostatic microinstabilities using a gyrokinetik model(S.Brunner, Ph.D thesis, Ecole Polytechnique Fédérale de Lausanne, Switzerland, thesis 1701 (1978)). By solving the spectral problem in a special Fourier space adapted to the curved geometry, orbit width as well as Larmor radius can be kept to all orders. For a first numerical implementation, a large aspect ratio plasma with circular concentric magnetic surfaces has been considered. A higher order Nyquist method(B.Davies, Jour.Comp.Phys. 66), 36 (1986), applied for identifying the eigenfrequencies in the complex plane, has been improved and enables straightforward implementation on a parallel computer. Illustrative results of ITG (ion temperature gradient) -related instabilities are presented. These include scaling studies of the radial width, toroidicity and magnetic shear scans, as well as the effects of non-adiabatic trapped electron dynamics.
Bounce-Transit and Drift Resonance and Neoclassical Toroidal Plasma Viscosity in Tokamaks
NASA Astrophysics Data System (ADS)
Shaing, K. C.; Chu, M. S.; Sabbagh, S. A.
2009-05-01
The importance of the resonance between the bounce frequency of the trapped particles and precession drift frequency in tokamaks to the low frequency magnetohydroynamic instabilities has been recognized for a long time. The resonance is also important in the transport processes as demonstrated by Park, et al. in calculating the neoclassical toroidal plasma viscosity [1]. They found that the transport fluxes are independent of the collision frequency, i.e., a resonant plateau regime. Here, we develop a theory for neoclassical toroidal plasma viscosity to include not only the bounce and drift resonance of the trapped particles but also the transit and drift resonance the circulating particles [2]. In the resonant plateau regime, our results are similar to those obtained by Park, et al., except that bounce average over the trapped particle trajectories is not performed and that the contributions from the circulating particles are included. In the collisional limit, it is found that the resonant plateau regime is connected to the Pfirsch-Schluter regime. [1] PARK, J.-K.,et al., IAEA,Fusion Energy Conference, Geneva, October 2008, Paper EX/5-3Rb. [2] SHAING, K. C., CHU, M. S., and SABBAGH, S. A., (to be submitted to Plasma Phys. Control. Fusion)
Wang, G. Q.; Ma, J.; Weiland, J.; Zang, Q.
2013-10-15
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.
Reid, R.L.; Barrett, R.J.; Brown, T.G.; Gorker, G.E.; Hooper, R.J.; Kalsi, S.S.; Metzler, D.H.; Peng, Y.K.M.; Roth, K.E.; Spampinato, P.T.
1985-03-01
The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged.
Multiscale gyrokinetics for rotating tokamak plasmas: fluctuations, transport and energy flows.
Abel, I G; Plunk, G G; Wang, E; Barnes, M; Cowley, S C; Dorland, W; Schekochihin, A A
2013-11-01
This paper presents a complete theoretical framework for studying turbulence and transport in rapidly rotating tokamak plasmas. The fundamental scale separations present in plasma turbulence are codified as an asymptotic expansion in the ratio ε = ρi/α of the gyroradius to the equilibrium scale length. Proceeding order by order in this expansion, a set of coupled multiscale equations is developed. They describe an instantaneous equilibrium, the fluctuations driven by gradients in the equilibrium quantities, and the transport-timescale evolution of mean profiles of these quantities driven by the interplay between the equilibrium and the fluctuations. The equilibrium distribution functions are local Maxwellians with each flux surface rotating toroidally as a rigid body. The magnetic equilibrium is obtained from the generalized Grad-Shafranov equation for a rotating plasma, determining the magnetic flux function from the mean pressure and velocity profiles of the plasma. The slow (resistive-timescale) evolution of the magnetic field is given by an evolution equation for the safety factor q. Large-scale deviations of the distribution function from a Maxwellian are given by neoclassical theory. The fluctuations are determined by the 'high-flow' gyrokinetic equation, from which we derive the governing principle for gyrokinetic turbulence in tokamaks: the conservation and local (in space) cascade of the free energy of the fluctuations (i.e. there is no turbulence spreading). Transport equations for the evolution of the mean density, temperature and flow velocity profiles are derived. These transport equations show how the neoclassical and fluctuating corrections to the equilibrium Maxwellian act back upon the mean profiles through fluxes and heating. The energy and entropy conservation laws for the mean profiles are derived from the transport equations. Total energy, thermal, kinetic and magnetic, is conserved and there is no net turbulent heating. Entropy is produced
Lafouti, Mansoureh; Ghoranneviss, Mahmood; Meshkani, Sakineh; Salar Elahi, Ahmad
2013-05-01
In this paper, both Resonant Helical magnetic Field (RHF) and limiter biasing have been applied to the tokamak. We have investigated their effects on the turbulence and transport of the particles at the edge of the plasma. The biased limiter voltage has been fixed at 200 V and RHF has L = 2 and L = 3. Also, the effects of the time order of the application of RHF and biasing to the tokamak have been explored. The experiment has been performed under three conditions. At first, the biasing and RHF were applied at t = 15 ms and at t = 20 ms. In the next step, RHF and biasing were applied at t = 15 ms and t = 20 ms, respectively. Finally, both of them were turned on at t = 15 ms until the end of the shot. For this purpose, the ion saturation current (I(sat)) and the floating potential (V(f)) have been measured by the Langmuir probe at r/a = 0.9. Moreover, the power spectra of I(sat) and floating potential gradient (∇V(f)), the coherency, the phase between them, and the particle diffusion coefficient have been calculated. The density fluctuations of the particles have been measured by the Rake probe and they have been analyzed with the Probability Distribution Function (PDF) technique. Also the particle diffusion coefficient has been determined by the Fick's law. The results show that, when RHF and biasing were applied at the same time to the plasma (during flatness region of plasma current), the radial particle density gradient, the radial particle flux, and the particle diffusion coefficient decrease about 50%, 60%, and 55%, respectively, compared to the other conditions. For more precision, the average values of the particle flux and the particle density gradient were calculated in the work. When the time is less than 15 ms, the average values of the particle flux and the particle density gradient are identical under all conditions, but in the other time interval they change. They reduce with the simultaneous application of biasing and RHF. The same results obtain
Ono, Masayuki; Furth, Harold
1993-01-01
An electron injection scheme for controlling transport in a tokamak plasma. Electrons with predominantly perpendicular energy are injected into a ripple field region created by a group of localized poloidal field bending magnets. The trapped electrons then grad-B drift vertically toward the plasma interior until they are detrapped, charging the plasma negative. Calculations indicate that the highly perpendicular velocity electrons can remain stable against kinetic instabilities in the regime of interest for tokamak experiments. The penetration distance can be controlled by controlling the "ripple mirror ratio", the energy of the injected electrons, and their v.sub..perp. /v.sub.51 ratio. In this scheme, the poloidal torque due to the injected radial current is taken by the magnets and not by the plasma. Injection is accomplished by the flat cathode containing an ECH cavity to pump electrons to high v.sub..perp..
NASA Astrophysics Data System (ADS)
Si, H.; Guo, H. Y.; Xu, G. S.; Xiao, B. J.; Luo, Z. P.; Guo, Y.; Wang, L.; Ding, R.
2016-03-01
Heat exhaust is one of the most challenging issues to be addressed for tokamak magnetic confinement fusion research. Detailed modeling with SOLPS5.0/B2.5-Eirene code package is carried out to examine an alternative advanced divertor configuration, i.e., quasi snowflake (QSF), for long pulse operation in EAST. Comparison is also made with the lower single null (LSN) divertor configuration. SOLPS predicts that the quasi snowflake configuration significantly reduces the peak heat flux at the lower divertor outer target, by a factor of 2-3, owing to the magnetic flux expansion. Furthermore, the density threshold for detachment is much lower for QSF, compared to LSN under the same upstream conditions. This indicates that QSF provides a promising tool for controlling heat flux at divertor target while maintaining a lower separatrix density, which is highly desirable for current drive, thus greatly facilitating long-pulse operation in EAST.
Sangwan, Deepak; Jha, Ratneshwar; Brotankova, Jana; Gopalkrishna, M. V.
2014-06-15
Parallel plasma flows in the scrape-off layer of ADITYA tokamak are measured in two orientations of total magnetic field. In each orientation, experiments are carried out by reversing the direction of the toroidal magnetic field and the plasma current. The transport-driven component is determined by averaging flow Mach numbers, measured in two directions of the toroidal magnetic field and the plasma current for the same orientation. It is observed that there is a significant transport-driven component in the measured flow and the component depends on the field orientation.
Lashkul, S. I.; Altukhov, A. B.; Gusakov, E. Z.; Dyachenko, V. V.; Esipov, L. A.; Irzak, M. A.; Kantor, M. Yu.; Kouprienko, D. V.; Saveliev, A. N.; Shatalin, S. V.; Stepanov, A. Yu.
2014-02-12
Results of comparative experimental studies of the efficiency of lower hybrid current drive (LHCD) and lower hybrid heating (LHH) in the FT-2 tokamak in hydrogen and deuterium plasmas are presented. In the new comparative experimental runs in deuterium/hydrogen plasmas suppression of the LHCD and beginning of the interaction of LH waves with ions is controlled by the plasma density rise. Role of parametric instabilities in CD switch-off is considered. In order to analyze the experimentally observed effect of LHCD the GRILL3D and FRTC codes has been used.
High-pressure duo-multichannel soft x-ray spectrometer for tokamak plasma diagnostics
Schwob, J.L.; Wouters, A.W.; Suckewer, S.
1987-03-01
A high-resolution, time-resolving soft X-ray multichannel spectrometer (SOXMOS) that permits the simultaneous measurement of emission in two different spectral ranges has been developed and tested extensively for tokamak plasma diagnostics. The basic instrument is a high-resolution, interferometrically adjusted, extreme grazing incidence Schwob-Fraenkel duochromator. The instrument is equipped with two multichannel detectors that are adjusted interferometrically and scan along the Rowland circle. Each consists of an MgF/sub 2/ coated, funneled microchannel plate, associated with a phosphor screen image intensifier that is coupled to a 1024-element photodiode array by a flexible fibrer optic conduit. The total wavelength coverage of the instrument is 5 to 340/sup 0/ A with a measured resolution (FWHM) of about 0.2 A when equipped with a 600 g/mm grating, and 5 to 85 A with a resolution of about 0.06 A using a 2400 g/mm grating. The simultaneous spectral coverage of each detector varies from 15 A at the short wavelength limit to 70 A at the long wavelength limit with the lower dispersion grating. The minimum read-out time for a full spectral portion is 17 ms, but several individual lines can be measured with 1 ms time resolution by selected pixel readout. Higher time resolution can be achieved by replacing one multichannel detector with a single channel electron multiplier detector. Examples of data from the PLT and TFTR tokamaks are presented to illustrate the instrument's versatility, high spectral resolution, and high signal-to-noise ratio even in the 10 A region. 44 refs., 20 figs.
NASA Astrophysics Data System (ADS)
Zanino, R.
1992-02-01
We have developed a 1 + 1 D time dependent code for the description of ion-impurity transport in a rotating tokamak plasma, using a pseudo-spectral discretization in the poloidal angle θ and a staggered finite difference mesh in the minor radius r. The plasma is assumed to have a constant uniform temperature T, to be in the high collisionality (Pfirsch-Schlüter) regime, and to contain electrons " e," fuel ions " i," and a single impurity species " Z" of charge eZ, where e is the proton charge. We are particularly interested in the case when: (1) flow velocities in the toroidal (symmetry) direction φ are in the range typical of neutral beam injection experiments, i.e., vthZ < Vφi, Z ⪅ vthi, ( vthj √2 T/ mj is the thermal speed, mj is the mass); (2) the relative concentration of impurities in the plasma, {ṅz}/{ṅi}, is significant and comparable to that observed in present tokamaks, i.e., √m e/m i ≪ ṅzZ 2/ ṅi ≈ 1 in order of magnitude. The model fluid equations are obtained via a moment approach, and an expansion in powers of the small ordering parameter δpi = ( mivthi/ eBθ) ((1/ ṅ | ∂ ṅi/∂r| ≪ 1 ( B is the magnetic field) is then employed. The equations at each order in δpi up to the second are solved, and the characteristic features of the results presented: to lowest order, outboard impurity peaking on each magnetic surface appears due to centrifugal forces; to first order, radial gradients driven ion-impurity friction gives rise to up-down asymmetries in the poloidal profiles; to second order, the radial profiles of density and rotation frequency evolve to steady state under the action of particle and angular momentum sources. The evolution of the poloidal profiles is decoupled from the evolution of the radial ones, thanks to the fact that the corresponding time scales belong to different orders in δpi: an algorithm is proposed to treat the 2D problem, alternating the solution of 1D problems.
NASA Astrophysics Data System (ADS)
Dimitrova, M.; Popov, Tsv K.; Ivanova, P.; Vasileva, E.; Hasan, E.; Horáček, J.; Vondráček, P.; Dejarnac, R.; Stöckel, J.; Weinzettl, V.; Havlicek, J.; Janky, F.; Panek, R.
2014-05-01
The scrape-off-layer (SOL) parameters in the COMPASS tokamak are studied by using a Langmuir probe mounted on a horizontal reciprocating manipulator. The radial profiles of the plasma potential, the electron energy distribution function and the electron densities are derived from the measured current-voltage probe characteristics by applying the firstderivative probe technique (FDPT). It is shown that close to the tokamak wall the electron energy distribution function is Maxwellian, while in the SOL, in the vicinity of the last closed flux surface and inside the confined plasma, the electron energy distribution function is bi-Maxwellian with a low-energy electron fraction dominating over a higher energy one. The radial profiles of the electron pressure and the parallel electron power flux density in COMPASS are also presented.
Kinetic full wave analyses of O-X-B mode conversion of EC waves in tokamak plasmas
NASA Astrophysics Data System (ADS)
Fukuyama, Atsushi; Khan, Shabbir Ahmad; Igami, Hiroe; Idei, Hiroshi
2016-10-01
For heating and current drive in a high-density plasma of tokamak, especially spherical tokamak, the use of electron Bernstein waves and the O-X-B mode conversion were proposed and experimental observations have been reported. In order to evaluate the power deposition profile and the current drive efficiency, kinetic full wave analysis using an integral form of dielectric tensor has been developed. The incident angle dependence of wave structure and O-X-B mode conversion efficiency is examined using one-dimensional analysis in the major radius direction. Two-dimensional analyses on the horizontal plane and the poloidal plane are also conducted, and the wave structure and the power deposition profile are compared with those of previous analyses using ray tracing method and cold plasma approximation. This work is supported by JSPS KAKENHI Grant Number JP26630471.
Core turbulent transport in tokamak plasmas: bridging theory and experiment with QuaLiKiz
NASA Astrophysics Data System (ADS)
Bourdelle, C.; Citrin, J.; Baiocchi, B.; Casati, A.; Cottier, P.; Garbet, X.; Imbeaux, F.; Contributors, JET
2016-01-01
Nonlinear gyrokinetic codes allow for detailed understanding of tokamak core turbulent transport. However, their computational demand precludes their use for predictive profile modeling. An alternative approach is required to bridge the gap between theoretical understanding and prediction of experiments. A quasilinear gyrokinetic model, QuaLiKiz (Bourdelle et al 2007 Phys. Plasmas 14 112501), is demonstrated to be rapid enough to ease systematic interface with experiments. The derivation and approximation of this approach are reviewed. The quasilinear approximation is proven valid over a wide range of core plasma parameters. Examples of profile prediction using QuaLiKiz coupled to the CRONOS integrated modeling code (Artaud et al 2010 Nucl. Fusion 50 043001) are presented. QuaLiKiz is being coupled to other integrated modeling platforms such as ETS and JETTO. QuaLiKiz quasilinear gyrokinetic turbulent heat, particle and angular momentum fluxes are available to all users. It allows for extensive stand-alone interpretative analysis and for first principle based integrated predictive modeling.
Plasma diagnostics in spherical tokamaks with silicon charged-particle detectors
NASA Astrophysics Data System (ADS)
Netepenko, A.; Boeglin, W. U.; Darrow, D. S.; Ellis, R.; Sibilia, M. J.
2016-11-01
Detection of charged fusion products, such as protons and tritons resulting from D(d, p) t reactions, can be used to determine the position and time dependent fusion reaction rate profile in spherical tokamak plasmas with neutral beam heating. We have developed a prototype instrument consisting of 6 ion-implanted-silicon surface barrier detectors combined with collimators in such a way that each detector can accept 3 MeV protons and 1 MeV tritons and thus provides a curved view across the plasma cross section. The combination of the results from all six detectors will provide information on the spatial distribution of the fusion reaction rate. The expected time resolution of about 1 ms makes it possible to study changes in the reaction rate due to slow variations in the neutral beam density profile, as well as rapid changes resulting from MHD instabilities. Details of the new instrument, its data acquisition system, simulation results, and electrical noise testing results are discussed in this paper. First experimental data are expected to be taken during the current experimental campaign at NSTX-U.
The effect of limiter conditioning on the Tokamak Fusion Test Reactor edge plasma
Kilpatrick, S.J.; Manos, D.M.; Nyberg, I.; Ramsey, A.T.; Stratton, B.C.; Timberlake, J.; Ulrickson, M.J.; Pitcher, C.S.; Dylla, H.F.
1991-07-01
Measurements by moveable Langmuir probes and edge spectroscopy diagnostics have documented the conditioning effect of low density helium-initiated discharge sequences on the Tokamak Fusion Test Reactor (TFTR) edge plasma. Langmuir probe measurements show in general that the edge electron density n{sub e} decreases by less than a factor of 2 while the edge electron temperature T{sub e} doubles. Radial profiles to the plasma boundary show that the density scrape-off length increases somewhat while the temperature scrape-off length decreases substantially. The particle flux density is unaffected. The spectral emission of C 2 decreases by a factor of 2, a much smaller change than that exhibited by the D{sub {alpha}} signal. These results complement previous accounts of the conditioning technique. Comparisons of these He conditioning measurements are made to edge measurements during a deuterium density scan experiment, showing many similarities, and to an existing edge model of the conditioning process, showing qualitative agreement. 20 refs., 5 figs.
The effect of limiter conditioning on the Tokamak Fusion Test Reactor edge plasma
Kilpatrick, S.J.; Pitcher, C.S.; Dylla, H.F.; Manos, D.M.; Nyberg, I.; Ramsey, A.T.; Stratton, B.C.; Timberlake, J.; Ulrickson, M.J. )
1991-05-01
Measurements by moveable Langmuir probes and edge spectroscopy diagnostics have documented the conditioning effect of low-density helium-initiated discharge sequences on the Tokamak Fusion Test Reactor edge plasma. Langmuir probe measurements shown in general that the edge electron density {ital n}{sub {ital e}} decreases by less than a factor of 2 while the edge electron temperature {ital T}{sub {ital e}} doubles. Radial profiles to the plasma boundary show that the density scrape-off length increases somewhat while the temperature scrape-off length decreases substantially. The particle flux density is unaffected. The spectral emission of C II decreases by a factor of 2, a much smaller change than that exhibited by the {ital D}{sub {alpha}} signal. These results complement previous accounts of the conditioning technique. Comparisons of these He conditioning measurements are made to edge measurements during a deuterium density scan experiment, showing many similarities, and to an existing edge model of the conditioning process, showing qualitative agreement.
Redd, A.J.; Kritz, A.H.; Bateman, G.; Horton, W.
1998-05-01
A drift wave transport model, recently developed by Ottaviani, Horton and Erba (OHE) [Ottaviani {ital et al.}, Plasma Phys. Controlled Fusion {bold 39}, 1461 (1997)], has been implemented and tested in a time-dependent predictive transport code. This OHE model assumes that anomalous transport is due to turbulence driven by ion temperature gradients and that the fully developed turbulence will extend into linearly stable regions, as described in the reference cited above. A multiplicative elongation factor is introduced in the OHE model and simulations are carried out for 12 discharges from major tokamak experiments, including both L- and H-modes (low- and high-confinement modes) and both circular and elongated discharges. Good agreement is found between the OHE model predictions and experiment. This OHE model is also used to describe the performance of the International Thermonuclear Experimental Reactor (ITER) [Putvinski {ital et al.}, in {ital Proceedings of the 16th IAEA Fusion Energy Conference}, Montr{acute e}al, Canada, 1996 (International Atomic Energy Agency, Vienna, 1997), Vol. 2, p. 737.] A second version of the OHE model, in which the turbulent transport is not allowed to penetrate into linearly stable regions, has also been implemented and tested. In simulations utilizing this version of the model, the linear stability of the plasma core eliminates the anomalous thermal transport near the magnetic axis, resulting in an increase in the core temperatures to well above the experimental values. {copyright} {ital 1998 American Institute of Physics.}
A Delta-f to Full-F PIC Simulation Scheme for Tokamak Plasmas
NASA Astrophysics Data System (ADS)
Lee, W. W.; Ethier, S.
2012-03-01
A generalized weight-based particle simulation schemes suitable for simulating microturbulence in magnetic fusion plasmas, where the zeroth-order inhomogeneity is important, has recently been developed [1]. The schemes is a generalization of the perturbative simulation schemes developed earlier for PIC simulations [2]. The new two-weight scheme, which can simulate both the perturbed distribution and the full distribution within the same code, has now been implemented to simulate tokamak plasmas using the GTC code [3]. Its development is based on the concept of multiscale expansion, which separates the scale lengths of the background inhomogeneity from those associated with the perturbed distributions. The code starts out as a delta-f code and gradually evolves into a full-F code, as such the delta-f part can help us with the noise issue in the linear stage and the full-F part can be useful in the fully nonlinear stage when the particle weights become too large or it becomes necessary to simulate realistic situations where sinks and sources become important.[4pt] [1] W. W. Lee, T. G. Jenkins and S. Ethier, Comp. Phys. Comm. 182, 564 (2011).[0pt] [2] S. E. Parker and W. W. Lee, Phys. Fluids B 5, 77 (1993).[0pt] [3] Z. Lin, T. S. Hahm, W. W. Lee, W. M. Tang and R. White, Science 281, 1835 (1998).
Plasma horizontal position control for the J-TEXT tokamak based on feedforward density compensation
NASA Astrophysics Data System (ADS)
Yu, W. Z.; Chen, Z. P.; Ke, X.; Li, F. M.; Zhu, L. Z.; Peng, Y. Y.; Zhang, M.; Zhuang, G.
2014-04-01
A relationship of the vertical field Bz required to maintain horizontal stability and the ratio of the central line-averaged density to the plasma current \\overline {n_{e}} /I_{p} is deduced and is proved to be linear through the previous statistical discharge data on the J-TEXT tokamak. According to this, a plasma horizontal position control system based on feedforward density compensation (FDC) is developed and tested on the machine for different ramping rates of density variation. Due to the feedforward density signal, the controller with FDC could suppress the disturbance of the horizontal displacement owing to the density variation during the flat-top phase moving more quickly when compared with the case with the traditional proportional-integral-derivative controller, which is a feedback controller. There is a trade-off between the suppression of the perturbation and the oscillation of the horizontal displacement for density variation, which should be taken into account in real routine discharge.
Optimized tomography methods for plasma emissivity reconstruction at the ASDEX Upgrade tokamak.
Odstrčil, T; Pütterich, T; Odstrčil, M; Gude, A; Igochine, V; Stroth, U
2016-12-01
The soft X-ray (SXR) emission provides valuable insight into processes happening inside of high-temperature plasmas. A standard method for deriving the local emissivity profiles of the plasma from the line-of-sight integrals measured by pinhole cameras is the tomographic inversion. Such an inversion is challenging due to its ill-conditioned nature and because the reconstructed profiles depend not only on the quality of the measurements but also on the inversion algorithm used. This paper provides a detailed description of several tomography algorithms, which solve the inversion problem of Tikhonov regularization with linear computational complexity in the number of basis functions. The feasibility of combining these methods with the minimum Fisher information regularization is demonstrated, and various statistical methods for the optimal choice of the regularization parameter are investigated with emphasis on their reliability and robustness. Finally, the accuracy and the capability of the methods are demonstrated by reconstructions of experimental SXR profiles, featuring poloidal asymmetric impurity distributions as measured at the ASDEX Upgrade tokamak.
NASA Astrophysics Data System (ADS)
Felici, Federico
2012-10-01
Recent experiments on TCV have demonstrated integrated control of the sawtooth and Neoclassical Tearing Mode (NTM) instabilities in a combined preemption-suppression strategy. This strategy is enabled by new sawtooth control methods (sawtooth pacing) in which modulation of sawtooth-stabilizing electron cyclotron power during the sawtooth cycle stimulates the advent of the crash. Rather than controlling the average sawtooth period, the precise timing of each individual crash can now be prescribed. Using this knowledge, efficient preemptive stabilization of NTMs becomes possible by applying power on the rational surface only at the instant of the crash-generating seed island. TCV experiments demonstrate that this approach, reinforced by NTM stabilization as a backup strategy, is effectively failsafe. This opens the road to inductive H-mode scenarios with long sawteeth providing longer inter-crash periods of high density and temperature. Also Edge Localized Modes are susceptible to EC modulation and it is shown that individual ELM events can be controlled using similar techniques. For advanced tokamak scenarios, MHD control is to be combined with optimization and control of the plasma kinetic and magnetic profile evolution in time. Real-time simulation of a physical model (RAPTOR) of current transport, including bootstrap current, neoclassical conductivity and auxiliary current drive, yields complete knowledge of the relevant profiles at any given time. The pilot implementation on TCV shows that these calculations can indeed be done in real-time and the resulting profiles have been included in feedback control schemes. Integration of this model with time-varying equilibria and internal current profile diagnostics provides a new framework for real-time interpretation of diagnostic data for plasma prediction, scenario monitoring, disruption prevention and feedback control.
NASA Astrophysics Data System (ADS)
Ren, Jing; Liu, Yueqiang; Liu, Yue; Medvedev, S. Yu; Wang, Zhirui; Xia, Guoliang
2016-11-01
The effects of an ideal/resistive conducting wall, the drift kinetic resonances, as well as the toroidal plasma flow, on the stability of the ideal external kink mode are numerically investigated for a reactor-relevant tokamak plasma with strongly negative triangularity (NTR) shaping. Comparison is made for a similar plasma equilibrium, but with positive triangularity (PTR). It is found that the ideal wall stabilization is less efficient for the kink stabilization in the NTR plasma due to a less ‘external’ eigenmode structure compared to the PTR plasma. The associated plasma displacement in the NTR plasma does not ‘balloon’ near the outboard mid-plane, as is normally the case for the pressure-driven kink-ballooning instability in PTR plasmas, but being more pronounced near the X-points. The toroidal flow plays a similar role for the kink stability for both NTR and PTR plasmas. The drift kinetic damping is less efficient for the ideal external kink mode in the NTR plasma, despite a somewhat larger fraction of the particle trapping near the plasma edge compared to the PTR equilibrium. However, the drift kinetic damping of the resistive wall mode (RWM) in the NTR plasma is generally as efficient as that of the PTR plasma, although the RWM window, in terms of the normalized pressure, is narrower for the NTR plasma.
NASA Astrophysics Data System (ADS)
Sen, WANG; Qiping, YUAN; Bingjia, XIAO
2017-03-01
Plasma control system (PCS), mainly developed for real-time feedback control calculation, plays a significant part during normal discharges in a magnetic fusion device, while the tokamak simulation code (TSC) is a nonlinear numerical model that studies the time evolution of an axisymmetric magnetized tokamak plasma. The motivation to combine these two codes for an integrated simulation is specified by the facts that the control system module in TSC is relatively simple compared to PCS, and meanwhile, newly-implemented control algorithms in PCS, before applied to experimental validations, require numerical validations against a tokamak plasma simulator that TSC can act as. In this paper, details of establishment of the integrated simulation framework between the EAST PCS and TSC are generically presented, and the poloidal power supply model and data acquisition model that have been implemented in this framework are described as well. In addition, the correctness of data interactions among the EAST PCS, Simulink and TSC is clearly confirmed during an interface test, and in a simulation test, the RZIP control scheme in the EAST PCS is numerically validated using this simulation platform. Supported by the National Magnetic Confinement Fusion Science Program of China (No. 2014GB103000) and the National Natural Science Foundation of China (No. 11205200).
NASA Astrophysics Data System (ADS)
Gilligan, John; Bourham, Mohamed
1993-09-01
Disruption damage conditions for future large tokamaks like ITER are nearly impossible to simulate on current tokamaks. The electrothermal plasma source SIRENS has been designed, constructed, and operated to produce high density (> 1025/m3), low temperature (1-3 eV) plasma formed by the ablation of the insulator with currents of up to 100 kA (100 μs pulse length) and energies up to 15 kJ. The source heat fluence (variable from 0.2 to 7 MJ/m2) is adequate for simulation of the thermal quench phase of plasma disruption in future fusion tokamaks. Different materials have been exposed to the high heat flux in SIRENS, where comparative erosion behavior was obtained. Vapor shield phenomena has been characterized for different materials, and the energy transmission factor through the shielding layer is obtained. The device is also equipped with a magnet capable of producing a parallel magnetic field (up to 16 T) over a 8 msec pulse length. The magnetic field is produced to decrease the turbulent energy transport through the vapor shield, which provides further reduction of surface erosion (magnetic vapor shield effect).
Gilligan, J.; Bourham, M. )
1993-09-01
Disruption damage conditions for future large tokamaks like ITER are nearly impossible to simulate on current tokamaks. The electrothermal plasma source SIRENS has been designed, constructed, and operated to produce high density (> 10[sup 25]/m[sup 3]), low temperature (1-3 eV) plasma formed by the ablation of the insulator with currents of up to 100 kA (100 [mu]s pulse length) and energies up to 15 kJ. The source heat fluence (variable from 0.2 to 7 MJ/m[sup 2]) is adequate for simulation of the thermal quench phase of plasma disruption in future fusion tokamaks. Different materials have been exposed to the high heat flux in SIRENS, where comparative erosion behavior was obtained. Vapor shield phenomena has been characterized for different materials, and the energy transmission factor through the shielding layer is obtained. The device is also equipped with a magnet capable of producing a parallel magnetic field (up to 16 T) over a 8 msec pulse length. The magnetic field is produced to decrease the turbulent energy transport through the vapor shield, which provides further reduction of surface erosion (magnetic vapor shield effect).
Magnetic confinement experiment. I: Tokamaks
Goldston, R.J.
1995-08-01
Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM`y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nT{tau}`s {approximately} 2.5x greater than ELM`ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices.
NASA Astrophysics Data System (ADS)
Todo, Y.
2016-11-01
Magnetohydrodynamic (MHD) instabilities driven by energetic particles in tokamak plasmas and the energetic particle distribution formed with the instabilities, neutral beam injection, and collisions are investigated with hybrid simulations for energetic particles and an MHD fluid. The multi-phase simulation, which is a combination of classical simulation and hybrid simulation, is applied to examine the distribution formation process in the collisional slowing-down time scale of energetic ions for various beam deposition power ({P}{NBI}) and slowing-down time ({τ }{{s}}). The physical parameters other than {P}{NBI} and {τ }{{s}} are similar to those of a Tokamak Fusion Test Reactor (TFTR) experiment (Wong et al 1991 Phys. Rev. Lett. 66 1874). For {P}{NBI} = 10 MW and {τ }{{s}} = 100 ms, which is similar to the TFTR experiment, the bursts of toroidal Alfvén eigenmodes take place with a time interval 2 ms, which is close to that observed in the experiment. The maximum radial velocity amplitude (v r) of the dominant TAE at the bursts in the simulation is {v}{{r}}/{v}{{A}}∼ 3× {10}-3 where v A is the Alfvén velocity at the plasma center. For {P}{NBI} = 5 MW and {τ }{{s}} = 20 ms, the amplitude of the dominant TAE is kept at a constant level {v}{{r}}/{v}{{A}}∼ 4× {10}-4. The intermittency of TAE rises with increasing {P}{NBI} and increasing {τ }{{s}} (= decreasing collision frequency). With increasing volume-averaged classical energetic ion pressure, which is well proportional to {P}{NBI}{τ }{{s}}, the energetic ion confinement degrades monotonically due to the transport by the instabilities. The volume-averaged energetic ion pressure depends only on the volume-averaged classical energetic ion pressure, not independently on {P}{NBI} or {τ }{{s}}. The energetic ion pressure profile resiliency, where the increase in energetic ion pressure profile is saturated, is found for the cases with the highest {P}{NBI}{τ }{{s}} where the TAE bursts take place.
NASA Astrophysics Data System (ADS)
Houyin, Wang; Jiansheng, Hu; Yaowei, Yu; Bin, Cao; Jinhua, Wu; Guoqing, Shen; Zhao, Wan; EAST, Contributors
2017-01-01
Although the deuterium and helium have almost the same mass, a Penning Optical Gas Analyzer (POGA) system on the basis of the spectroscopic method and Penning discharging has been designed on EAST, since 2014. The POGA system was developed successfully in 2015, it was the first time that EAST could detect helium partial pressure in deuterium plasma (wall conditioning and plasma operation scenario). With dedicated calibration and proper adjustment of the parameters, the minimum concentration of helium in deuterium gas can be measured as about 0.5% instead of 1% on the other tokamak devices. Moreover, the He and D2 partial pressures are measured simultaneously. At present, the measurable range of deuterium partial pressure is 1 × 10-7 mbar to 1 × 10-5 mbar, meanwhile the range of helium is 1 × 10-8 mbar to 1 × 10-5 mbar. The measurable range can be modified by means of the adjustment of POGA system’s parameters. It is possible to detect the interesting part of the gas with a time resolution of less than 5 ms (the 200 ms because of conductance of transfer pipe at present). The POGA system was routinely employed to wall conditioning and helium enrichment investigation in 2015. Last but not the least, the low temperature plasma of POGA is generated by normal penning gauge Pfeiffer IKR gauge instead of Alcatel CF2P, which has been suspended for a few years and was used for almost all the POGA systems in the world.
Characterization of Plasma Gun with TiH2/C60 Cartridge for Disruption Mitigation in Tokamaks
NASA Astrophysics Data System (ADS)
Bogatu, I. N.; Thompson, J. R.; Galkin, S. A.; Kim, J. S.; HyperV Technologies Corp. Team
2011-10-01
Impurity injection for disruption mitigation in tokamaks must be faster than growth time of plasma instabilities, requires sufficient mass to get critical electron density, high penetrability, and large assimilation fraction in the core plasma, with rapid impurity redistribution over the whole plasma. FAR-TECH, Inc. proposed the innovative idea to use hyper-velocity (>30 km/s), high-density (>1023 m-3) C60/C plasma jets with high ram pressure to deliver the impurity mass in <1 ms. For this purpose C60 powder explosively sublimated into molecular gas, from a solid state, pulsed power driven TiH2/C60 injector cartridge is ionized and accelerated in a plasma accelerator. We report the complete characterization of the TiH2/C60 cartridge with 5 kJ capacitive driver which demonstrated the capability of producing >30 mg of C60 gas in <0.5 ms. In addition we present the construction and testing status of a 100 kJ coaxial plasma gun (~35 cm length) prototype with TiH2/C60 cartridge for a small scale, proof-of-principle experiment on a tokamak. Work supported by the US DOE DE-FG02-08ER85196 grant.
Disruption avoidance in the SINP-Tokamak by means of electrode-biasing at the plasma edge
NASA Astrophysics Data System (ADS)
Basu, Debjyoti; Pal, Rabindranath; Martinell, Julio J.; Ghosh, Joydeep; Chattopadhyay, Prabal K.
2013-05-01
Control of plasma disruption by a biased edge electrode is reported in SINP-Tokamak. The features that characterize a plasma disruption are reduced with increasing bias potential. The disruption can be completely suppressed with the concomitant stabilization of observed MHD modes that are allegedly precursors of the disruption. An m = 3/n = 1 tearing mode, which apparently causes disruption can be stabilized when a negative biasing potential is applied near the edge. These changes in the disruptive behavior with edge biasing are hypothesized to be due to changes in the current density profile.
Resistive reduced MHD modeling of multi-edge-localized-mode cycles in Tokamak X-point plasmas.
Orain, F; Bécoulet, M; Huijsmans, G T A; Dif-Pradalier, G; Hoelzl, M; Morales, J; Garbet, X; Nardon, E; Pamela, S; Passeron, C; Latu, G; Fil, A; Cahyna, P
2015-01-23
The full dynamics of a multi-edge-localized-mode (ELM) cycle is modeled for the first time in realistic tokamak X-point geometry with the nonlinear reduced MHD code jorek. The diamagnetic rotation is found to be instrumental to stabilize the plasma after an ELM crash and to model the cyclic reconstruction and collapse of the plasma pressure profile. ELM relaxations are cyclically initiated each time the pedestal gradient crosses a triggering threshold. Diamagnetic drifts are also found to yield a near-symmetric ELM power deposition on the inner and outer divertor target plates, consistent with experimental measurements.
NASA Astrophysics Data System (ADS)
Fitzpatrick, Richard
2016-05-01
The effect of the perturbed ion polarization current on the stability of neoclassical tearing modes in tokamak plasmas is calculated using an improved, neoclassical, four-field, drift-magnetohydrodynamical model. The calculation involves the self-consistent determination of the pressure and scalar electric potential profiles in the vicinity of the associated magnetic island chain, which allows the chain's propagation velocity to be fixed. Two regimes are considered. First, a regime in which neoclassical ion poloidal flow damping is not strong enough to enhance the magnitude of the polarization current (relative to that found in slab geometry). Second, a regime in which neoclassical ion poloidal flow damping is strong enough to significantly enhance the magnitude of the polarization current. In both regimes, two types of solution are considered. First, a freely rotating solution (i.e., an island chain that is not interacting with a static, resonant, magnetic perturbation). Second, a locked solution (i.e., an island chain that has been brought to rest in the laboratory frame via interaction with a static, resonant, magnetic perturbation). In all cases, the polarization current is found to be either always stabilizing or stabilizing provided that ηi≡d ln Ti/d ln ne does not exceed some threshold value. In certain ranges of ηi, the polarization current is found to have a stabilizing effect on a freely rotating island, but a destabilizing effect on a corresponding locked island.
Hybrid simulation of energetic particle effects on tearing modes in tokamak plasmas
Cai Huishan; Fu Guoyong
2012-07-15
The effects of energetic ions on stability of tearing mode are investigated by global kinetic/MHD hybrid simulations in a low beta tokamak plasma. The kinetic effects of counter circulating energetic ions from the non-adiabatic response are found to be strongly destabilizing while the effects from the adiabatic response are stabilizing. The net effect with both adiabatic and non-adiabatic contributions is destabilizing. On the other hand, the kinetic effects of co-circulating energetic ions from the non-adiabatic response are calculated to be weakly stabilizing while the corresponding adiabatic contribution is destabilizing for small energetic ion beta. The net effect is weakly stabilizing. The dependence of kinetic effects on energetic ion beta, gyroradius, and speed is studied systematically and the results agree in large part with the previous analytic results for the kinetic effects of circulating particles. For trapped energetic ions, their effects on tearing mode stability are dominated by the adiabatic response due to large banana orbit width and strong poloidal variation of particle pressure. The net effect of trapped energetic particles on tearing modes is much more destabilizing as compared to that of counter circulating particles at the same beta value.
Self-Consistent Simulation of Turbulence and Transport in Tokamak Edge Plasmas
Rognlien, T D; Umansky, M V; Xu, X Q; Cohen, R H
2003-09-03
The status of coupling the fluid 3D turbulence code BOUT and the fluid plasma/neutral 2D transport code UEDGE is reported, where both codes simulate the edge region of diverted tokamaks from several cm inside the magnetic separatrix to the far scrape-off layer (SOL), thereby including the magnetic X-point. Because the characteristic time scale of the turbulence is short ({approx} 10{sup -5}-10{sup -4}s) and the profile evolution time scale can be long ({approx} 10{sup -2}-10{sup -1} s owing to recycling), an iterative scheme is used that relaxes the turbulent fluxes passed from BOUT to UEDGE and the profiles from UEDGE to BOUT over many coupling steps. Each code is run on its own characteristic time scale, yielding a statistically averaged steady state. For this initial study, the ion and neutral densities and parallel velocities are evolved, while the temperature profiles are stationary. Here the turbulence code is run in the electrostatic approximation. For this example of self-consistent coupling with strong L-mode-like turbulence, the ion flux to the main-chamber exceeds that to the divertor plates.
Lu, Z. X.; Tynan, G.; Wang, W. X.; Ethier, S.; Diamond, P. H.; Gao, C.; Rice, J.
2015-05-15
Intrinsic torque, which can be generated by turbulent stresses, can induce toroidal rotation in a tokamak plasma at rest without direct momentum injection. Reversals in intrinsic torque have been inferred from the observation of toroidal velocity changes in recent lower hybrid current drive (LHCD) experiments. This work focuses on understanding the cause of LHCD-induced intrinsic torque reversal using gyrokinetic simulations and theoretical analyses. A new mechanism for the intrinsic torque reversal linked to magnetic shear (s{sup ^}) effects on the turbulence spectrum is identified. This reversal is a consequence of the ballooning structure at weak s{sup ^}. Based on realistic profiles from the Alcator C-Mod LHCD experiments, simulations demonstrate that the intrinsic torque reverses for weak s{sup ^} discharges and that the value of s{sup ^}{sub crit} is consistent with the experimental results s{sup ^}{sub crit}{sup exp}≈0.2∼0.3 [Rice et al., Phys. Rev. Lett. 111, 125003 (2013)]. The consideration of this intrinsic torque feature in our work is important for the understanding of rotation profile generation at weak s{sup ^} and its consequent impact on macro-instability stabilization and micro-turbulence reduction, which is crucial for ITER. It is also relevant to internal transport barrier formation at negative or weakly positive s{sup ^}.
Lu, Z. X.; Wang, W. X.; Diamond, P. H.; Tynan, G.; Ethier, S.; Gao, C.; Rice, J.
2015-05-04
We report that intrinsic torque, which can be generated by turbulent stresses, can induce toroidal rotation in a tokamak plasma at rest without direct momentum injection. Reversals in intrinsic torque have been inferred from the observation of toroidal velocity changes in recent lower hybrid current drive (LHCD) experiments. Here we focus on understanding the cause of LHCD-induced intrinsic torque reversal using gyrokinetic simulations and theoretical analyses. A new mechanism for the intrinsic torque reversal linked to magnetic shear (sˆ) effects on the turbulence spectrum is identified. This reversal is a consequence of the ballooning structure at weak sˆ . Based on realistic profiles from the Alcator C-Mod LHCD experiments, simulations demonstrate that the intrinsic torque reverses for weak sˆ discharges and that the value of sˆ _{crit} is consistent with the experimental results sˆ ^{exp}_{crit} [Rice et al., Phys. Rev. Lett. 111, 125003 (2013)]. In conclusion, the consideration of this intrinsic torque feature in our work is important for the understanding of rotation profile generation at weak and its consequent impact on macro-instability stabilization and micro-turbulence reduction, which is crucial for ITER. It is also relevant to internal transport barrier formation at negative or weakly positive sˆ .
On extended analytic theory of 2D ballooning modes in tokamak plasmas
NASA Astrophysics Data System (ADS)
Abdoul, Peshwaz; Dickinson, David; Roach, Colin; Wilson, Howard
2016-10-01
We have extended the leading order ballooning theory which typically yields more unstable isolated mode (IM) that usually sit on the outboard mid-plane, to higher order where less unstable general mode (GM) sits at a different poloidal location. Our analytic theory has revealed that any poloidal shift of the mode with respect to the outboard mid-plane - arising from the effect of profile variations, for example - is always accompanied by an asymmetry of the radial eigenmode structure. Hence, GMs have radial asymmetry. Our theory can have important consequences, especially for calculations that invoke quasilinear theory to model intrinsic rotation arising from Reynolds stress. This is very important in ITER for which external torques are small. In such theories it is the radial asymmetry in the global GM mode which can generate a Reynolds stress that could in principle contribute to the poloidal flow during the low to high (L-H) mode transition in tokamaks. I am also an associate member at the York Plasma Institute, University of York and teaching at the Physics Department, University of Sulaimani, Kurdistan Region, Iraq.
Energetic particle driven geodesic acoustic mode in a toroidally rotating tokamak plasma
NASA Astrophysics Data System (ADS)
Ren, Haijun
2017-01-01
Energetic particle (EP) driven geodesic acoustic modes (EGAMs) in toroidally rotating tokamak plasmas are analytically investigated using the hybrid kinetic-fluid model and gyrokinetic equations. By ignoring high-order terms and ion Landau damping, the kinetic dispersion relation is reduced to the hybrid one in the large safety factor limit. There is one high-frequency branch with a frequency larger than {ωt0} , the transit frequency of EPs with initial energy, which is always stable. Two low-frequency solutions with a frequency smaller than {ωt0} are complex conjugates in the hybrid limit. In the presence of ion Landau damping, the growth rate of the unstable branch is decreased and the damping rate of the damped branch is increased. The toroidal Mach number is shown to increase {{ Ω }\\text{r}} , the normalized real frequency of both branches. Although not affecting the instability critical condition, the Mach number decreases the growth rate when {{ Ω }\\text{r}} is larger than a critical value Ω \\text{r}\\text{cri} and enlarges the growth rate when {{ Ω }\\text{r}}< Ω \\text{r}\\text{cri} . The ion Landau damping effect is negligible for large M. But the discrepancy between the kinetic dispersion relation and the hybrid one becomes ignorable only for q≳ 7 .
Simulation of a tokamak edge plasma with the kinetic code COGENT
NASA Astrophysics Data System (ADS)
Dorf, M.; Cohen, R.; Dorr, M.; Hittinger, J.; Rognlien, T.; Colella, P.; Martin, D.; McCorquodale, P.
2013-10-01
Progress on the development of the continuum gyrokinetic code COGENT for edge plasma simulations is reported. The COGENT code models an axisymmetric gyrokinetic equation coupled to the long-wavelength limit of the gyro-Poisson equation. COGENT is distinguished by application of fourth-order conservative discretization, and mapped multiblock grid technology to handle the geometric complexity of the tokamak edge. The code has also a number of model collision operator options, which have been successfully verified in neoclassical simulations. Our recent development work has focused on incorporation of the full (nonlinear) Fokker-Planck collision model. The implementation of the Fokker-Plank operator is discussed in detail, and the results of the initial verification studies are presented. In addition, we report on progress and status of the newly available divertor version of the COGENT code that includes both closed and open magnetic field line regions and a model for recycled neutral gas. Work performed for USDOE, at LLNL under contract DE-AC52-07NA27344 and at LBNL under contract DE-AC02-05CH11231.
Long-lasting energetic particle modes in tokamak plasmas with low magnetic shear
NASA Astrophysics Data System (ADS)
Zhang, Rui-Bin; Wang, Xian-Qu; Xiao, Chi-Jie; Wang, Xiao-Gang; Liu, Yi; Deng, Wei; Chen, Wei; Ding, Xuan-Tong; Duan, Xu-Ru; the HL-2 A Team
2014-09-01
A long-lasting (for hundreds of milliseconds) m/n = 1 energetic particle mode driven by trapped fast ions, other than conventional fishbone bursts, is studied theoretically and in comparison with HL-2A experimental results. The mode can be observed in weak shear tokamak plasmas during neutral beam injection with a mostly steady amplitude envelope of long-lasting magnetic perturbation signals. The dispersion relation and radial structure of the mode are calculated with a weak shear q-profile. Both the m/n = 1/1 component and its higher frequency m/n = 2/2 harmonics are found to be unstable, in good agreement with experimental observations on HL-2A. On the other hand, due to the feature of weak magnetic shear, the mode is also significantly different from bursty fishbones, especially the mode structure, temporal behavior, instability threshold and growth rate dependence on the fast ion gradient. The nonlinear evolution of the mode and the comparison with fishbone bursts are also further investigated.
Pigarov, A Y; West, W; Soukhanovskii, V; Rognlien, T; Maingi, R; Lipschultz, B; Krasheninnikov, S; LaBombard, B
2003-11-25
Fast intermittent transport has been observed in the scrape-off layer (SOL) of major tokamaks including Alcator C-Mod, DIII-D, and NSTX. This kind of transport is not diffusive but rather convective. It strongly increases plasma flux to the chamber walls and enhances the recycling of neutral particles in the main chamber. We discuss anomalous cross-field convection (ACFC) model for impurity and main plasma ions and its relation to intermittent transport events, i.e. plasma density blobs and holes in the SOL. Along with plasma diffusivity coefficients, our transport model introduces time-independent anomalous cross-field convective velocity. In the discharge modelling, diffusivity coefficients and ACFC velocity profiles are adjusted to match a set of representative experimental data. We use this model in the edge plasma physics code UEDGE to simulate the multi-fluid two-dimensional transport for these three tokamaks. We present simulation results suggesting the dominance of anomalous convection in the far SOL transport. These results are consistent with the hypothesis that the chamber wall is an important source of impurities and that different impurity charge states have different directions of anomalous convective velocity.
Stacey, W. M.; Bae, C.
2015-06-15
A systematic formalism for the calculation of rotation in non-axisymmetric tokamaks with 3D magnetic fields is described. The Braginskii Ωτ-ordered viscous stress tensor formalism, generalized to accommodate non-axisymmetric 3D magnetic fields in general toroidal flux surface geometry, and the resulting fluid moment equations provide a systematic formalism for the calculation of toroidal and poloidal rotation and radial ion flow in tokamaks in the presence of various non-axisymmetric “neoclassical toroidal viscosity” mechanisms. The relation among rotation velocities, radial ion particle flux, ion orbit loss, and radial electric field is discussed, and the possibility of controlling these quantities by producing externally controllable toroidal and/or poloidal currents in the edge plasma for this purpose is suggested for future investigation.
Effect of deuteron temperature on iron forbidden line intensities in rf-heated tokamak plasmas
Sato, K.; Suckewer, S.; Wouters, A.
1987-05-01
Two line ratios, the forbidden line at 845.5 A (2s/sup 2/2p /sup 2/P/sub 1/2/ - 2s/sup 2/2p /sup 2/P/sub 3/2/) to the allowed line at 135.7 A (2s/sup 2/2p /sup 2/P/sub 1/2/ - 2s2p/sup 2/ /sup 2/D/sub 3/2/) in Fe XXII and the forbidden line at 592.1 A (2s/sup 2/2p/sup 4/ /sup 3/P/sub 2/ - 2s/sup 2/2p/sup 4/ /sup 1/D/sub 2/) to the forbidden line at 1118.2 A (2s/sup 2/2p/sup 4/ /sup 3/P/sub 2/ - 2s/sup 2/2p/sup 4/ /sup 3/P/sub 1/) in Fe XIX, have been measured as the ion temperature-sensitive line ratios during rf heating in the Princeton Large Torus. The results indicate that deuteron collisions in plasmas of high deuteron temperature have a noticeable effect on the intensity of the forbidden lines. Measured relative intensities are compared with values from level population calculations, which include deuteron collisional excitation between the levels of the ground configuration. The agreement between the observed and calculated ratios is within 30%. A method for deuteron (or proton) temperature measurement in tokamak plasmas is discussed. 37 refs.
A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma
Ku, S.; Hager, R.; Chang, C.S.; Kwon, J.M.; Parker, S.E.
2016-06-15
In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles provide scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation – e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others – can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function – driven by ionization, charge exchange and wall loss – is allowed to be arbitrarily large. The numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.
A new hybrid-Lagrangian numerical scheme for gyrokinetic simulation of tokamak edge plasma
Ku, S.; Hager, R.; Chang, C. S.; Kwon, J. M.; Parker, S. E.
2016-06-01
In order to enable kinetic simulation of non-thermal edge plasmas at a reduced computational cost, a new hybrid-Lagrangian δf scheme has been developed that utilizes the phase space grid in addition to the usual marker particles, taking advantage of the computational strengths from both sides. The new scheme splits the particle distribution function of a kinetic equation into two parts. Marker particles contain the fast space-time varying, δf, part of the distribution function and the coarse-grained phase-space grid contains the slow space-time varying part. The coarse-grained phase-space grid reduces the memory-requirement and the computing cost, while the marker particles provide scalable computing ability for the fine-grained physics. Weights of the marker particles are determined by a direct weight evolution equation instead of the differential form weight evolution equations that the conventional delta-f schemes use. The particle weight can be slowly transferred to the phase space grid, thereby reducing the growth of the particle weights. The non-Lagrangian part of the kinetic equation – e.g., collision operation, ionization, charge exchange, heat-source, radiative cooling, and others – can be operated directly on the phase space grid. Deviation of the particle distribution function on the velocity grid from a Maxwellian distribution function – driven by ionization, charge exchange and wall loss – is allowed to be arbitrarily large. The numerical scheme is implemented in the gyrokinetic particle code XGC1, which specializes in simulating the tokamak edge plasma that crosses the magnetic separatrix and is in contact with the material wall.
Modified neural networks for rapid recovery of tokamak plasma parameters for real time control
NASA Astrophysics Data System (ADS)
Sengupta, A.; Ranjan, P.
2002-07-01
Two modified neural network techniques are used for the identification of the equilibrium plasma parameters of the Superconducting Steady State Tokamak I from external magnetic measurements. This is expected to ultimately assist in a real time plasma control. As different from the conventional network structure where a single network with the optimum number of processing elements calculates the outputs, a multinetwork system connected in parallel does the calculations here in one of the methods. This network is called the double neural network. The accuracy of the recovered parameters is clearly more than the conventional network. The other type of neural network used here is based on the statistical function parametrization combined with a neural network. The principal component transformation removes linear dependences from the measurements and a dimensional reduction process reduces the dimensionality of the input space. This reduced and transformed input set, rather than the entire set, is fed into the neural network input. This is known as the principal component transformation-based neural network. The accuracy of the recovered parameters in the latter type of modified network is found to be a further improvement over the accuracy of the double neural network. This result differs from that obtained in an earlier work where the double neural network showed better performance. The conventional network and the function parametrization methods have also been used for comparison. The conventional network has been used for an optimization of the set of magnetic diagnostics. The effective set of sensors, as assessed by this network, are compared with the principal component based network. Fault tolerance of the neural networks has been tested. The double neural network showed the maximum resistance to faults in the diagnostics, while the principal component based network performed poorly. Finally the processing times of the methods have been compared. The double
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.
Lu, B.; Wang, F.; Fu, J.; Li, Y.; Wan, B.; Shi, Y.; Bitter, M.; Hill, K. W.; Lee, S. G.
2012-10-15
Two imaging x-ray crystal spectrometers, the so-called 'poloidal' and 'tangential' spectrometers, were recently implemented on experimental advanced superconducting tokamak (EAST) to provide spatially and temporally resolved impurity ion temperature (T{sub i}), electron temperature (T{sub e}) and rotation velocity profiles. They are derived from Doppler width of W line for Ti, the intensity ratio of Li-like satellites to W line for Te, and Doppler shift of W line for rotation. Each spectrometer originally consisted of a spherically curved crystal and a two-dimensional multi-wire proportional counter (MWPC) detector. Both spectrometers have now been upgraded. The layout of the tangential spectrometer was modified, since it had to be moved to a different port, and the spectrometer was equipped with two high count rate Pilatus detectors (Model 100 K) to overcome the count rate limitation of the MWPC and to improve its time resolution. The poloidal spectrometer was equipped with two spherically bent crystals to record the spectra of He-like and H-like argon simultaneously and side by side on the original MWPC. These upgrades are described, and new results from the latest EAST experimental campaign are presented.
Zang, Qing; Zhao, Junyu; Chen, Hui; Li, Fengjuan; Hsieh, C. L.
2013-09-15
The detector circuit is the core component of filter polychromator which is used for scattering light analysis in Thomson scattering diagnostic, and is responsible for the precision and stability of a system. High signal-to-noise and stability are primary requirements for the diagnostic. Recently, an upgraded detector circuit for weak light detecting in Experimental Advanced Superconducting Tokamak (EAST) edge Thomson scattering system has been designed, which can be used for the measurement of large electron temperature (T{sub e}) gradient and low electron density (n{sub e}). In this new circuit, a thermoelectric-cooled avalanche photodiode with the aid circuit is involved for increasing stability and enhancing signal-to-noise ratio (SNR), especially the circuit will never be influenced by ambient temperature. These features are expected to improve the accuracy of EAST Thomson diagnostic dramatically. Related mechanical construction of the circuit is redesigned as well for heat-sinking and installation. All parameters are optimized, and SNR is dramatically improved. The number of minimum detectable photons is only 10.
Goedbloed, J. P.
2012-06-15
It is shown that some of the main results of the recent paper by Lakhin and Ilgisonis [Phys. Plasmas 18, 092103 (2011)], viz. the derivation of the equations for the continuous spectra of poloidally and toroidally rotating plasmas and their special solution for large aspect ratio tokamaks with large parallel flows were obtained before by Goedbloed, Belieen, van der Holst, and Keppens [Phys. Plasmas 11, 28 (2004)]. A further rearrangement of the system of equations for the coupled Alfven and slow continuous spectra clearly exhibits: (a) coupling through a single tangential derivative, which is a generalization of the geodesic curvature; (b) the 'transonic' transitions of the equilibrium, which need to be carefully examined in order to avoid entering hyperbolic flow regimes where the stability formalism breaks down. A critical discussion is devoted to the implications of this failure, which is generally missed in the tokamak literature, possibly as a result of the wide-spread use of the sonic Mach number of gas dynamics, which is an irrelevant and misleading parameter in 'transonic' magnetohydrodynamics. Once this obstacle in understanding is removed, further application of the theory of trans-slow Alfven continuum instabilities to both tokamaks, with possible implications for the L-H transition, and astrophysical objects like 'fat' accretion disks, with a possible new route to magnetohydrodynamic turbulence, becomes feasible.
X-ray line emission from highly ionised argon and sulphur in tokamak plasmas
NASA Astrophysics Data System (ADS)
McGinnity, Paul
Observations of H-like and He-like argon line emission and associated satellite spectra have been made on the JET (Joint European Torus) tokamak by a Bragg rotorspectrometer and a double crystal monochromator. Similar He-like sulphur measurements have been made on the COMPASS-D (Compact Assembly) tokamak by a Johann curved crystal spectrometer. Recently calculated electron impact excitation rates for He-like ions were used in the derivation of the electron temperature sensitive line ratio G=(Ix+Iy+Iz)/Iw and the electron density sensitive ratio R=Iz/(Ix+Iy), where w, x, y and z are the He-like resonance line, intercombination lines and forbidden line respectively. For S XV the ratios Ik/Iw and Iq/Iw were also calculated, where k and q are Li-like dielectronic satellites to the w line formed by dielectronic recombination and inner shell excitation respectively. Both are electron temperature dependent, the latter also being sensitive to changes in the ionisation balance. The fine structure ratios Ix/Iy and /beta = Ly/alpha 2/Lyα1 were calculated for He-like S XV and H-like Ar XVIII respectively, where Ly/alpha/sb[1,2] are the fine structure components of the H-like Lyman alpha line. Transport modelling was carried out to account for non-coronal conditions in the JET plasma while a near-coronal equilibrium was assumed in the COMPASS-D plasma. Calculated ratios were compared with experimental measurements obtained from JET and COMPASS-D. For higher temperatures, such as during additional heating, the Ar XVII emission shell was found to move of axis, with a subsequent reduction in the G ratio. For S XV good agreement with calculations was found between the measured G and Iq/Iw ratios, indicating that the assumption of near-coronal equilibrium was valid. Lower than expected values of the S XV R ratio were found. After investigation of the atomic physics processes it was concluded that this was due to an unidentified instrumental effect of the Johann spectrometer. An
NASA Astrophysics Data System (ADS)
Kouprienko, D. V.; Altukhov, A. B.; Gurchenko, A. D.; Gusakov, E. Z.; Kantor, M. Yu.; Lashkul, S. I.; Esipov, L. A.
2010-05-01
The dynamics of electron heat transport at improved energy confinement during lower hybrid plasma heating in the FT-2 tokamak was studied experimentally. Evolution of the profiles of the electron temperature and density was thoroughly investigated under conditions of fast variation in the plasma parameters. The energy balance in the electron channel is calculated with the help of the ASTRA code by using the measured plasma parameters. Correlation is revealed between the dynamics of electron heat transport and the behavior of small-scale drift turbulence measured using the enhanced scattering correlation diagnostics. The suppression of heat transfer and turbulence agrees well with the increase in the shear of poloidal plasma rotation calculated from experimental data in the neoclassical approximation.
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.
Murray, J.G.; Gorker, G.E.
1985-01-01
Tokamak fusion reactors will have large plasma currents of approximately 10 MA with hundreds of megajoules stored in the magnetic fields. When a major plasma instability occurs, the disruption of the plasma current induces voltage in the adjacent conducting structures, giving rise to large transient currents. The induced voltages may be sufficiently high to cause arcing across sector gaps or from one protruding component to another. This report reviews a tokamak arcing scenario and provides guidelines for designing tokamaks to minimize the possibility of arc damage.
Rognlien, T; Umansky, M; Xu, X; Cohen, R; LoDestro, L
2004-05-24
The edge-plasma profiles and fluxes to the divertor and walls of a divertor tokamak with a magnetic X-point are simulated by coupling a 2D transport code (UEDGE) and a 3D turbulence code (BOUT). An relaxed iterative coupling scheme is used where each code is run on its characteristic time scale, resulting in a statistical steady state. Plasma variables of density, parallel velocity, and separate ion and electron temperatures are included, together with a fluid neutral model for recycling neutrals at material surfaces. Results for the DIII-D tokamak parameters show that the turbulence is preferentially excited in the outer radial region of the edge where magnetic curvature is destabilizing and that substantial plasma particle flux is transported to the main chamber walls. These results are qualitatively consistent with some experimental observations. The coupled transport/turbulence simulation technique provides a strategy to understanding edge-plasma physics in more detailed than previously available and to significantly enhance the realism of predictions of the performance of future devices.