Counter-neutral Beam Injection Research Articles

Sawtooth activities in EAST neutral beam injection plasma

Neutral beam injection (NBI) system has been proved to affect sawtooth activities through both producing energetic particles and supplying torque applied to the plasma. The impact of NBI on sawtooth crashes has been studied in the EAST tokamak, which is equipped with co- and counter NBIs. Statistical analysis shows that both strong co- and counter-NBI yield stronger sawtooth activities than cases when heating power is weak. A minimum sawtooth period is observed at a counter-NBI power of 0.2 MW. This is linked to zero-plasma rotation and a non-rotating precursor mode of the sawtooth instability. This indicates that the sawtooth instability is stabilized by plasma rotation as previously suggested. The fast-ion content differs between co- and counter-NBI, so energetic particles might additionally contribute to the sawtooth activities.

Turbulent particle transport as a function of toroidal rotation in DIII-D H-mode plasmas

In this paper we show how changes in toroidal rotation, by controlling the injected torque, affect particle transport and confinement. The toroidal rotation is altered using the co- and counter neutral beam injection (NBI) in low collisionality H-mode plasmas on DIII-D (Luxon 2002 Nucl. Fusion 42 614) with dominant electron cyclotron heating (ECH). We find that there is no correlation between the toroidal rotation shear and the inverse density gradient, which is observed on AUG when is varied using ECH (Angioni et al 2011 Phys. Rev. Lett. 107 215003). In DIII-D, we find that in a discharge with balanced torque injection, the shear is smaller than the linear gyrokinetic growth rate for small for –0.85. This results in lower particle confinement. In the co- and counter- injected discharges the shear is larger or close to the linear growth rate at the plasma edge and both configurations have higher particle confinement. In order to measure particle transport, we use a small periodic perturbative gas puff. This gas puff perturbs the density profiles and allows us to extract the perturbed diffusion and inward pinch coefficients. We observe a strong increase in the inward particle pinch in the counter-torque injected plasma. Finally, the calculated quasi-linear particle flux, nor the linear growth rates using TGLF (Staebler et al 2005 Phys. Plasmas 12 102508) agree with experimental observations.

Boundary condition for toroidal plasma flow imposed at the separatrix in high confinement JT-60U plasmas with edge localized modes and the physics process in pedestal structure formation

Modulation charge eXchange recombination spectroscopy measurements with high spatial and temporal resolution have made the evaluation of the toroidal plasma flow of fully stripped carbon impurity ions (VϕC6+) in the JT-60U tokamak peripheral region (including, in particular, the separatrix) possible with a better signal-to-noise ratio. By comparing co- and counter-neutral beam injection discharges experimentally, we have identified the boundary condition of VϕC6+ and radial electric field shear (∇Er) imposed at the separatrix in high confinement (H-mode) plasmas with edge localized modes (ELMs). The VϕC6+ value at the separatrix is not fixed at zero but varies with the momentum input direction. On the other hand, the ∇Er value is nearly zero (or very weakly positive) at the separatrix. Furthermore, the edge localized mode perturbation does not appear to affect both VϕC6+ and ∇Er values at the separatrix as strongly as that in the pedestal region. The above experimental findings based on the precise edge measurements have been used to validate a theoretical model and develop a new empirical model. A better understanding of the physical process in the edge transport barrier (ETB) formation due to the sheared Er formation is also discussed.

Stabilization of energetic-ion-driven MHD modes by ECCD in Heliotron J

Second harmonic electron cyclotron current drive (ECCD) has been applied in the stellarator/heliotron (S/H) device, Heliotron J, to stabilize magnetohydrodynamic (MHD) modes. The energetic particle mode (EPM) of 60–90 kHz frequency, one of the energetic-ion-driven MHD modes, is excited in a plasma heated by co- and counter-neutral beam injection and electron cyclotron heating (ECH). The EPM has been stabilized by counter-ECCD which decreases the rotational transform. Localized EC current driven by a few kA at the central region modifies the rotational transform profile, ι/2π, leading to the formation of a high magnetic shear at the radius where the mode is excited. An experiment scanning the EC-driven current shows that there is a threshold in magnetic shear and/or rotational transform to stabilize the EPM.

Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak Using Neutral Beam Injection

We model the internal transport barrier “ITB” in edge plasma of small size divertor tokamak with B2SOLPS0.5.2D fluid transport code. The simulation results demonstrated the following: (1) we control the internal transport barrier by altering the edge particle transport through changes the edge toroidal rotation which agree with the result of Burrell et al. (Edge Pedestal control in quiescent H-mode discharges in DIII-D using co-plus counter-neutral beam injection, Nucl Fusion, 49, 085024 (9pp) in 2009). (2) The radial electric field has neoclassical nature near separatrix with discharge by co-injection NBI. (3) The toroidal plasma viscosity has strong influence on the toroidal velocity.

Formation of Electron Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak

The formation of electron internal transport barrier (EITB) during using counter-neutral beam injection (NBI) heating in the edge plasma of small size divertor tokamak can be simulated by using fluid transport code B2SOLPS0.5.2D. The results of simulations give us the following: (1) Plasma heating with counter-neutral beam injection leads to, strong, parabola type electron internal transport barrier (EITB) was formed in the edge plasma of small size divertor tokamak. (2) In case of plasma heating by counter-neutral beam injection, the radial electric field shear (Er–gradient) was increased, while electron transport coefficients were reduced in conjunction with the formation of electron internal transport barrier (EITB). (3) The plasma heating by counter-neutral beam injection play significantly role in redistribution of parallel (toroidal) velocity in edge plasma of small size divertor tokamak.

Improvement of Plasma Core Confinement Via Electron-Root Realization by Strongly Focused ECRH in LHD: Core Electron-Root Confinement

Core electron-root confinement (CERC), observed in the Large Helical Device as well as in other helical devices, is an improved electron energy confinement mode. It is characterized by a highly peaked electron temperature profile in the core region and appears when the centrally focused electron cyclotron resonance heating power exceeds a certain threshold value. This threshold value has been clarified to associate with the transition of the radial electric field (Er) from the ion root (small negative value) to the electron root (large positive value greater than a few kV/m), based on the bifurcation nature of Er due to the ambipolarity condition of neoclassical transport fluxes that is specific in nonaxisymmetric configurations. It has been experimentally recognized that a steeper Te gradient is realized with a clear transition (power threshold nature) in target plasmas with counter neutral beam injection (NBI) than ones with codirectional NBI. It has been interpreted, based on the heat pulse propagation experiment, to be related to the rational surface or the island induced by the NBI-driven current. Transport analyses have shown that the incremental thermal diffusivity of electron heat transport becomes lower, and the standard thermal diffusivity decreases with the increase of heating power in CERC plasmas.

DIII-D research in support of ITER

DIII-D research is providing key information for the design and operation of ITER. Investigations of axisymmetric stability and of edge-localized mode (ELM) suppression with resonant magnetic perturbations have helped provide the physics basis for new axisymmetric and non-axisymmetric control coils in ITER. Discharges that simulate ITER operating scenarios in conventional H-mode, advanced inductive, hybrid and steady state regimes have achieved normalized performance consistent with ITER's goals for fusion performance. Stationary discharges with high βN and 90% non-inductive current that project to Q = 5 in ITER have been sustained for a current relaxation time (∼2.5 s), and high beta wall-stabilized discharges with fully non-inductive current drive have been sustained for more than one second. Detailed issues of plasma control have been addressed, including the development of a new large-bore startup scenario for ITER. DIII-D research also contributes to the basis for reliable operation in ITER, through active control of the chief performance-limiting instabilities. Simultaneous stabilization of neoclassical tearing modes (by localized current drive) and resistive wall modes (by magnetic feedback) has allowed stable operation at high beta and low rotation. In research aimed at improving the lifetime of material surfaces near the plasma, recent experiments have investigated several approaches to mitigation of disruptions, including injection of low-Z gas and low-Z pellets, and have shown the conditions that minimize core impurity accumulation during radiative divertor operation. Investigation of carbon erosion, transport and co-deposition with hydrogenic species, and methods for the removal of co-deposits, will contribute to the physics basis for initial operation of ITER with a carbon divertor. A broad research programme provides the physics basis for predicting the performance of ITER. Recent key results include the discovery that the L–H power threshold is reduced with low neutral beam torque, and the development of a successful model for prediction of the H-mode pedestal height in DIII-D. Research areas with the potential to improve ITER's performance include the demonstration of ELM-free ‘quiescent H-mode’ discharges with both co- and counter-neutral beam injection, and validation of the predicted torque generated by static, non-axisymmetric magnetic fields. New diagnostics provide detailed benchmarking of turbulent transport codes and direct measurements of the anomalous transport of fast ions by Alfvén instabilities. Successful comparison of experiment and modelling for off-axis neutral beam current drive provides the basis for more flexible current profile control in advanced scenarios.

Edge pedestal control in quiescent H-mode discharges in DIII-D using co-plus counter-neutral beam injection

We have made two significant discoveries in our recent studies of quiescent H-mode (QH-mode) plasmas in DIII-D. First, we have found that we can control the edge pedestal density and pressure by altering the edge particle transport through changes in the edge toroidal rotation. This allows us to adjust the edge operating point to be close to, but below the ELM stability boundary, maintaining the ELM-free state while allowing up to a factor of two increase in edge pressure. The ELM boundary is significantly higher in more strongly shaped plasmas, which broadens the operating space available for QH-mode and leads to improved core performance. Second, for the first time on any tokamak, we have created QH-mode plasmas with strong edge co-rotation; previous QH-modes in all tokamaks had edge counter-rotation. This result demonstrates that counter-NBI and edge-counter rotation are not essential conditions for QH-mode. Both these investigations benefited from the edge stability predictions based on peeling–ballooning mode theory. The broadening of the ELM-stable region with plasma shaping is predicted by that theory. The theory has also been extended to provide a model for the edge harmonic oscillation that enhances edge transport in the QH-mode. Many of the features of that theory agree with the experimental results reported either previously or in this paper. One notable example is the prediction that co-rotating QH-mode is possible provided sufficient shear in the edge rotation can be created.

Tearing mode physics studies applying the dynamic ergodic divertor on TEXTOR

The dynamic ergodic divertor (DED) on the TEXTOR tokamak allows for the reproducible destabilization of the m/n = 2/1 tearing mode which is phase locked to the external static or rotating perturbation field. In combination with its flexible heating systems (co- and counter-neutral beam injection, ion cyclotron resonance heating, electron cyclotron resonance heating (ECRH) with steerable launcher) dedicated experiments to study the mode onset, properties of large islands and mode stabilization can be performed. The dependence of the mode excitation threshold (field penetration) on the plasma rotation shows a resonance character, with minimum threshold when the external perturbation frequency matches the MHD frequency of the 2/1 mode. Mode stabilization by ECRH heating shows that for the TEXTOR plasma heating is more effective than the current drive in O-point. Extrapolation to ITER yields a significant contribution to the mode suppression originating from the temperature increase within the island. Alfvén-like modes, which have been previously identified in the vicinity of large islands on FTU (Buratti et al 2005 Nuclear Fusion 45 1446), are found to be created already before island formation above a certain threshold of the externally applied perturbation field.

Progress toward fully noninductive, high beta conditions in DIII-D

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario.

Studies of Radial Electric Field Formation and Its Effect on Fluctuations and Transport Barriers in the JFT-2M Tokamak

The mechanism of E × B flow formation, the effect of the E × B flow on parallel flow, the reduction of fluctuations by the shearing effect of the E × B flow shear, and the relation between the geodesic acoustic mode (GAM) and density fluctuations are discussed based on the experiments using various Er measurements in the JFT-2M tokamak. The experiments in plasmas with H-mode and counter-neutral beam injection (NBI) mode show that the feedback loop of the E × B flow shear, the fluctuation suppression, and an increase of ion diamagnetic flow are key to the formation of the transport barrier in toroidal plasmas. Two important effects of the radial electric field are presented: One is fluctuation suppression by the E × B flow shear, and the other is a drive of the parallel flow by radial electric field, which explains the driving mechanism of a spontaneous toroidal flow. The relation between the GAM and the density fluctuations is also discussed. The GAM is observed to be excited by the nonlinear coupling of density fluctuations, while the GAM itself affects the amplitude of the density fluctuations.

Sawtooth control in fusion plasmas

Clear observations of early triggering of neo-classical tearing modes by sawteeth with long quiescent periods have motivated recent efforts to control, and in particular destabilize, sawteeth. One successful approach explored in TCV utilizes electron cyclotron heating in order to locally increase the current penetration time in the core. The latter is also achieved in various machines by depositing electron cyclotron current drive or ion cyclotron current drive close to the q = 1 rational surface. Crucially, localized current drive also succeeds in destabilizing sawteeth which are otherwise stabilized by a co-existing population of energetic trapped ions in the core. In addition, a recent reversed toroidal field campaign at JET demonstrates that counter-neutral beam injection (NBI) results in shorter sawtooth periods than in the Ohmic regime. The clear dependence of the sawtooth period on the NBI heating power and the direction of injection also manifests itself in terms of the toroidal plasma rotation, which consequently requires consideration in the theoretical interpretation of the experiments. Another feature of NBI, expected to be especially evident in the negative ion based neutral beam injection (NNBI) heating planned for ITER, is the parallel velocity asymmetry of the fast ion population. It is predicted that a finite orbit effect of asymmetrically distributed circulating ions could strongly modify sawtooth stability. Furthermore, NNBI driven current with non-monotonic profile could significantly slow down the evolution of the safety factor in the core, thereby delaying sawteeth.

Heat pulse propagation across the rational surface in a large helical device plasma with counter-neutral beam injection

Core electron temperature profile flattening is observed in a large helical device [A. Iiyoshi et al. Nucl. Fusion 39, 1245 (1999)] inward shifted plasma with counter-neutral beam injection. To study this phenomenon, heat pulse experiments are performed by on-axis electron cyclotron heating power modulation. A unique feature of heat pulse propagation is observed near the m∕n=2∕1 rational surface (m, n are the poloidal and toroidal mode numbers, respectively). A simultaneous response of the temperature perturbation on radially separated flux surfaces is shown. The change in the magnetic field topology due to the presence of a magnetic island structure can explain this nonmonotonic heat pulse propagation. The estimated O-point position of the island is located near the m∕n=2∕1 rational surface.

Studies of the ‘Quiescent H-mode’ regime in ASDEX Upgrade and JET

The stationary edge localized mode (ELM)-free ‘Quiescent H-mode’ (QH-mode) regime, obtained with counter neutral beam injection, is studied in ASDEX Upgrade and Joint European Torus. QH-mode plasmas have high pedestal and core ion temperatures together with good High-confinement mode (H-mode). ELMs are replaced by continuous MHD oscillations, the ‘edge harmonic oscillation’ (EHO) and the ‘high frequency oscillation’. Stationarity of particle and impurity densities is linked to the occurrence of these MHD modes. The EHO location in the steep-gradient region and its appearance with increasing edge pressure points towards the edge pressure or pressure gradient as possible drivers for the EHO. Injection of small cryogenic pellets can raise the plasma density to about 40% of the Greenwald density limit without triggering ELMs. Orbit following calculations of the slowing-down distribution show the presence of an enhanced fast particle density in the H-mode barrier region despite the large loss currents with counter-injection. The radial electrical field in the edge barrier region, about twice as large in QH-mode as in ELMy H-mode, is large enough to reverse the precession drift direction of injected beam ions, leading to a resonance of the EHO with the drift precession frequency.

Plasma rotation profile measurements using Doppler reflectometry

High spatial resolution radial profiles of the perpendicular plasma rotation velocity u⊥ using a dual channel 50–75 GHz Doppler reflectometer system on the ASDEX Upgrade tokamak are presented for a variety of discharge scenarios, including Ohmic, L-mode, H-mode, etc with forward and reversed magnetic field and co- and counter neutral beam injection. The reflectometers have steppable launch frequencies fo = c/λo, with selectable O- or X-mode polarization, giving tokamak edge to mid-radius coverage. Low-field-side antennae (hog-horn antenna pairs) with deliberate tilting (primarily poloidally) produce a Doppler shifted spectrum directly proportional to the perpendicular velocity fD = u⊥k⊥/2π = u⊥ 2sinθt/λo. The incident angle θt between the beam and cut-off layer normal varies with plasma shape, cut-off layer position and refraction. However, typical angles range from 5° to 27° giving a probed turbulence wavenumber, k⊥, range of 1.8–14.3 cm−1, with resulting Doppler shifts fD of up to 5 MHz. The measured perpendicular velocity is u⊥ = vE × B + vphase, which for a typical H-mode is slightly positive in the tokamak scrape-off-layer with a deep negative well across the H-mode steep pressure gradient pedestal region and then following the perpendicularly projected toroidal fluid velocity in the core, should be dominated by the E × B velocity, as the intrinsic phase velocity is predicted to be small, which may allow u⊥ to be interpreted directly as the radial electric field Er profile.

Edge ion parameters at the L–H transition on JET

L–H transition experiments have recently been performed on JET to characterize the local edge ion parameters. Evidence of a threshold in the edge ion temperature to access the H-mode is presented and comparisons are made with edge electron temperature. These results provide the first scaling of pedestal ion temperature on JET. The threshold edge ion temperature scales approximately linearly with toroidal magnetic field and shows no dependence on electron density in the range explored. Reference shots have been repeated with the ion ∇ B drift away from the X-point and counter neutral beam injection. The L–H transition is also considered in terms of dimensionless parameter space. The effective edge ion collisionality does not appear to play a significant role in the L–H transition, providing no clear distinction between the L- and H-modes. In contrast, the normalized poloidal ion gyroradius provides a sharp H-mode boundary at any given field and current, consistent with a threshold ion temperature for H-mode access being dependent on toroidal field but not on density. However, comparison of ρ* and with forward and reversed fields across a range of fields and currents demonstrates that these dimensionless parameters simply describe an operation trajectory for the JET L–H transition.

Dynamics of Plasma during the Formation of a Weak Negative Central Magnetic Shear Configuration Using Counter Neutral Beam Injection in the JFT-2M Tokamak.

In the JFT-2M tokamak, discharges exhibiting good performance have been produced by counterNBI after boronization. An improved core confinement mode under an H-mode edge condition is sustained for -5~~E With H89p - 1.5 and pN - 2.0 at a line-averaged electron density of around 70%-80% of the Greenwald density limit. It is found that the q-profile changes from a monotonic one having qo ~ 1 to a zero or weak negative central magnetic shear configuration having I < qmin ~ qo < 3 during improved core confinement mode using counter-NBI under an H-mode edge condition. The central electron temperature very clearly increases when the sawtooth activity disappears, while the previous result without boronization exhibited a significant degradation of the energy confinement due to the accumulation of impurities. The mechanism governing the improved core confinement mode including the density peaking and/or ITB formation due to negative shear is presently unclear.

Understanding and control of transport in Advanced Tokamak regimes in DIII-D

Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection taking place when the heating power exceeds a threshold value of about 9 MW, contrasting to co-NBI injection, where Pthreshold&amp;lt;2.5 MW. Expansion of the ITB is enhanced compared to similar co-injected discharges. Both differences are believed to arise from modification of the E×B shear dynamics when the sign of the rotation contribution is reversed. Sustainment of an AT regime with βNH89=9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode (edge localized, high confinement mode) edge and an ITB, and exhibiting ion thermal transport down to 2–3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency magnetohydrodynamic (MHD) activity is identified as the probable cause.

Modification by means of counter-neutral beam injection in a low- plasma

The idea of controlling the radial electric field by means of a radial current resulting from ion orbit loss caused by counter neutral beam injection has been theoretically and experimentally investigated. A large fraction (%) of the 75 keV deuterium ions counter-injected into a low- plasma ( MA) suffers prompt orbit loss, which forces an inward ion current to maintain charge neutrality. Monte Carlo guiding-centre orbit calculations predict a radial current of 80 A at the last closed flux surface. In these discharges, is negative everywhere, owing to the counter-going toroidal rotation, and exhibits a double-bump shape, in contrast to the usual positive parabolic shape for the co-injection case. The measured carbon impurity ion toroidal rotation profile shows a pedestal over the outer region where fast ions are lost, possibly due to the effect of torque. The momentum diffusion process tends to slow down and to spatially spread the torque effect. The L - H transition did not occur more quickly in these discharges than in similar co-injected discharges.