Off-axis Neutral Beam Injection Research Articles

Gyrokinetic analysis of turbulent transport by electromagnetic turbulence in finite β plasmas with weak magnetic shear on HL-2A

Turbulent transport and zonal flow (ZF) dynamics in the mixed kinetic ballooning mode (KBM) and ion temperature gradient (ITG) mode dominated turbulence states are analyzed by gyrokinetic simulations based on the experimental observations of KBMs in the HL-2A Internal Transport Barrier (ITB) plasmas with weak magnetic shear configuration. Results have shown that KBMs and ITGs are the main candidates for turbulent transport and the inclusion of fast ions (FIs) can destabilize both instabilities, which is resulted by the decrease of ballooning parameter αMHD hence reduced Shafranov shift stabilization due to the negative FI density gradient, making them more unstable according to the ŝ-αMHD diagram. In addition, the ITGs are suppressed by both dilution and finite-β stabilization effects caused by FIs. Nonlinear simulations excluding the effects of FIs indicate that the transport is minimum at βe=βeexp as the turbulence predominated by ITGs is strongly suppressed by the nonlinear electromagnetic (EM) stabilization and enhancement of ZFs. However, the transport is further increased with β e although the ZFs become stronger due to the transition from ITG to KBM dominated turbulence state. The presence of FIs can modify the relation between ZF shearing rate and β e. The transport level is insensitive to β e when βe⩽βeexp but increases significantly because of the KBM destabilization. Meanwhile, the ZF amplitude reaches maximum at βe=βeexp whereas it suffers an erosion at higher values, implying that the plasma might be a self-organized critical state owning to the interactions among ZFs, KBMs and FI effects hence setting the plasma β around β exp. The results have also provided a possible explanation that the destabilization of EM turbulence is responsible for the ion heat transport stiffness under weak magnetic shear configurations in the off-axis neutral beam injection heated ITB plasmas on HL-2A.

Influence of Off-Axis Neutral Beam Injection on Resistive Wall Mode Stability

Investigated are the effects of energetic particles (EPs), arising from off-axis neutral beam injection, on the resistive wall mode (RWM) instability in tokamaks. The analytic results for a cylindrical plasma show that EPs contribute the strongest kinetic damping to the RWM, as the deposition radius approaches the peak location of the mode eigenfunction. Toroidal computations are also carried out utilizing the MARS-K code, which includes EPs with anisotropic equilibrium distribution in the particle pitch angle space. These self-consistent numerical results confirm the analytic finding, that the RWM growth rate decreases as the peak of the EPs’ radial density profile approaches the maximum of the mode displacement. This study thus demonstrates the possibility of the RWM stabilization by the off-axis neutral beam injection in tokamak experiments.

Observation of Alfvén Eigenmodes driven by off-axis neutral beam injection in the TCV tokamak

Fast-particle driven Alfvén Eigenmodes have been observed in low-collisionality discharges with off-axis neutral beam injection (NBI), electron cyclotron resonance heating (ECRH) and a reduced toroidal magnetic field. During NBI and ECRH, toroidicity induced Alfvén Eigenmodes (TAEs) appear in frequency bands close to 200 kHz and energetic-particle-induced geodesic acoustic modes (EGAMs) are observed at about 40 and 80 kHz. When turning off ECRH in the experiment, those beam-driven modes disappear which can be explained by a modification of the fast-ion slowing down distribution. In contrast, coherent fluctuations close to the frequency of the beam-driven TAEs are present throughout the experiment. The modes are even observed during ohmic plasma conditions, which clearly demonstrates that they are not caused by fast particles and suggests an alternative drive, such as turbulence. The mode-induced fast-ion transport has been found to be weak and marginal in terms of the fast-ion diagnostic sensitivities. Measurements of the plasma stored energy, neutron rates, neutral particle fluxes and fast-ion D-alpha spectroscopy show good agreement with neoclassical modelling results from TRANSP. This is further supported by a similarly good agreement between measurement and modelling in cases with and without ECRH and therefore with and without the modes. Instead, a significant level of charge exchange losses are predicted and observed which generate a bump-on-tail fast-ion distribution function that can provide the necessary free energy to EGAMs.

Saturated helical mode in EAST high β hybrid plasmas

A hybrid scenario with the minimum safety factor above unity and a safety factor profile with a wide range of low magnetic shear in the core is one of the candidates for ITER operations. Theory predicts that such equilibria are prone to quasi-interchange mode, which was confirmed by observations from several devices. The presence of those 3D quasi-interchange helical modes has both positive and negative effects on plasma performance. On the one hand, they can avoid low mode number instabilities such as sawteeth. On the other hand, they would lead to confinement degradation, significant fast ion losses and toroidal rotation damping. Such modes are observed for the first time in EAST high βp hybrid plasmas achieved with off-axis neutral beam injection (NBI) heating. The mode manifested itself as long-lived saturated helical instability. Notably, when the mode was destabilized, the toroidal rotation profile in the core was completely flattened. The extent of the flattened rotation profile reaches up to half of the minor radius. As the minimum safety factor qmin approaches unity, strong n = 2 and n = 3 harmonics are successively driven unstable as nonlinear consequences of the n = 1 mode. A 3D resistive nonlinear simulation with a realistic tokamak configuration has been applied to reproduce the evolution of long-lived helical instability with the M3D code. The simulated magnetic perturbation shows a strong n = 2 component. As the mode grows, the amplitude of the n = 2 harmonic grows immensely with respect to that of the n = 1 harmonic. The n = 2 harmonic gradually becomes dominant as the mode saturates. In saturation, the initial weak shear safety factor profile in the core becomes completely flat. This change of safety factor profile in the core corroborates a self-regulating magnetic flux pumping mechanism which is considered responsible for maintaining stationary non-sawtoothing hybrid discharges.

An edge fast-ion D-alpha system installed at ASDEX Upgrade

An edge fast-ion D-alpha (FIDA) system consisting of lines of sight optimized for edge measurements and a newly designed prototype spectrometer have been installed at ASDEX Upgrade. The system allows for measurements of the full deuterium Balmer alpha spectrum with either good radial coverage or exposure times below 200 μs. It uses a bar, placed in front of the electron multiplying charge-coupled device camera, to partially block the intense unshifted D-alpha light which permits simultaneous measurements of the FIDA and unshifted D-alpha emission. Thus, not only active FIDA radiation present during neutral beam injection (NBI) can be studied but also detailed investigations of passive FIDA light are possible, since the unshifted D-alpha emission contains information on the density of background neutrals. First measurements show significant levels of active and passive FIDA radiation in discharges with large edge fast-ion populations from off-axis NBI heating. In addition, a particularly strong response of the passive emission to edge localized modes is observed, well explained by modifications of the edge-neutral density as well as the fast-ion density.

Numerical simulations of global Alfvén eigenmodes excitation and stabilization in NSTX-U

Three-dimensional nonlinear simulations of Alfvén eigenmodes in the subcyclotron frequency range show a robust physical stabilizing mechanism via modest off-axis beam injection, in agreement with experimental observations from the National Spherical Torus Experiment (NSTX-U). Experimental results from NSTX-U have demonstrated that neutral beam injection from the new beam sources with large tangency radii deposits beam ions with large pitch, which can very effectively stabilize all unstable Global Alfvén Eigenmodes (GAEs). Beam-driven GAEs have been linked to enhanced electron transport in NSTX, and the ability to control these modes will have significant implications for NSTX-U, ITER, and other fusion devices where super-Alfvénic fast ions might be present. Nonlinear simulations using the HYM code have been performed to study the excitation and stabilization of GAEs in the NSTX-U right before and shortly after the additional off-axis beam injection. The simulations reproduce the experimental finding, namely, it is shown that off-axis neutral beam injection reliably and strongly suppresses all unstable GAEs. Before additional beam injection, the simulations show unstable counter-rotating GAEs with toroidal mode numbers and frequencies that match the experimentally observed modes. Additional off-axis beam injection has been modeled by adding beam ions with large pitch and varying density. The complete stabilization occurs at less than 7% of the total beam ion inventory. New analytical theory of GAE (de)stabilization has also been derived, suggesting a different interpretation for the GAE stabilization mechanism compared to previous publications.

The neutron camera upgrade for MAST Upgrade

The Neutron Camera Upgrade (NCU) is a neutron flux monitor consisting of six lines of sight (LoSs) under installation on Mega Ampere Spherical Tokamak (MAST) Upgrade. The NCU is expected to contribute to the study of the confinement of fast ions and on the efficiency of non-inductive current drive in the presence of on-axis and off-axis neutral beam injection by measuring the neutron emissivity profile along the equatorial plane. This paper discusses the NCU main design criteria, the engineering and interfacing issues, and the solutions adopted. In addition, the results from the characterization and performance studies of the neutron detectors using standard γ-rays sources and a 252Cf source are discussed. The proposed design has a time resolution of 1 ms with a statistical uncertainty of less than 10% for all MAST Upgrade scenarios with a spatial resolution of 10 cm: higher spatial resolution is possible by moving the LoSs in-between plasma discharges. The energy resolution of the neutron detector is better than 10% for a light output of 0.8 MeVee, and the measured pulse shape discrimination is satisfactory.

Overview of the present progress and activities on the CFETR

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

Fishbone Modes in Plasmas with Dual Neutral Beam Injection Heating

The density gradient of fast ions is the main driving force for fishbone instability that in turn results in fast ion loss. It is possible to reduce the instability by eliminating the density gradient of the fast ions by employing dual neutral beam injection (DNBI) in tokamak plasmas. The dispersion relation for the fishbone instability is applied to the case of DNBI with suitable fast ion distribution functions. The results show that the density distribution of fast ions of DNBI can bring about a stable window that is a range of values for the distance between the on-axis beam and the off-axis beam that yields an overall stabilization of the resultant fishbone mode. The width of the stable window increases linearly with the position of the safety factor q = 1 magnetic flux surface increasing. In addition, the width of the stable window becomes wider for a more peaked density profile of fast ions and keeps constant for a peaked enough density profile of fast ions. The growth rates of the fishbone modes dramatically decrease with the intensity ratio of off-axis neutral beam injection (NBI) and on-axis NBI, and the critical beta values of fast ions increase with the intensity ratio increasing. Fishbone modes can be avoided with DNBI, which may be an effective method to prevent fast ion loss resulting from fishbone instabilities.

The high-βN hybrid scenario for ITER and FNSF steady-state missionsa)

New experiments on DIII-D have demonstrated the steady-state potential of the hybrid scenario, with 1 MA of plasma current driven fully non-inductively and βN up to 3.7 sustained for ∼3 s (∼1.5 current diffusion time, τR, in DIII-D), providing the basis for an attractive option for steady-state operation in ITER and FNSF. Excellent confinement is achieved (H98y2 ∼ 1.6) without performance limiting tearing modes. The hybrid regime overcomes the need for off-axis current drive efficiency, taking advantage of poloidal magnetic flux pumping that is believed to be the result of a saturated 3/2 tearing mode. This allows for efficient current drive close to the axis, without deleterious sawtooth instabilities. In these experiments, the edge surface loop voltage is driven down to zero for >1 τR when the poloidal β is increased above 1.9 at a plasma current of 1.0 MA and the ECH power is increased to 3.2 MW. Stationary operation of hybrid plasmas with all on-axis current drive is sustained at pressures slightly above the ideal no-wall limit, while the calculated ideal with-wall MHD limit is βN ∼ 4–4.5. Off-axis Neutral Beam Injection (NBI) power has been used to broaden the pressure and current profiles in this scenario, seeking to take advantage of higher predicted kink stability limits and lower values of the tearing stability index Δ′, as calculated by the DCON and PEST3 codes. Results based on measured profiles predict ideal limits at βN > 4.5, 10% higher than the cases with on-axis NBI. A 0-D model, based on the present confinement, βN and shape values of the DIII-D hybrid scenario, shows that these plasmas are consistent with the ITER 9 MA, Q = 5 mission and the FNSF 6.7 MA scenario with Q = 3.5. With collisionality and edge safety factor values comparable to those envisioned for ITER and FNSF, the high-βN hybrid represents an attractive high performance option for the steady-state missions of these devices.

Combined magnetic and kinetic control of advanced tokamak steady state scenarios based on semi-empirical modelling

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.

Quantification of the impact of large and small-scale instabilities on the fast-ion confinement in ASDEX Upgrade

The confinement fast ions, generated by neutral beam injection (NBI), has been investigated at the ASDEX Upgrade tokamak. In plasmas that exhibit strong sawtooth crashes, a significant sawtooth-induced internal redistribution of mainly passing fast ions is observed, which is in very good agreement with the theoretical predictions based on the Kadomtsev model. Between the sawtooth crashes, the fishbone modes are excited which, however, do not cause measurable changes in the global fast-ion population. During experiments with on- and off-axis NBI and without strong magnetohydrodynamic (MHD) modes, the fast-ion measurements agree very well with the neo-classical predictions. This shows that the MHD-induced (large-scale), as well as a possible turbulence-induced (small-scale) fast-ion transport is negligible under these conditions. However, in discharges performed to study the off-axis NBI current drive efficiency with up to 10 MW of heating power, the fast-ion measurements agree best with the theoretical predictions that assume a weak level anomalous fast-ion transport. This is also in agreement with measurements of the internal inductance, a Motional Stark Effect diagnostic and a novel polarimetry diagnostic: the fast-ion driven current profile is clearly modified when changing the NBI injection geometry and the measurements agree best with the predictions that assume weak anomalous fast-ion diffusion.

Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis of the pedestal highlights the potential roles of micro-tearing modes and kinetic ballooning modes for the pedestal formation. Mitigation of edge localized modes (ELM) using resonant magnetic perturbation has been demonstrated for toroidal mode numbers n = 3, 4, 6 with an ELM frequency increase by up to a factor of 9, compatible with pellet fuelling. The peak heat flux of mitigated and natural ELMs follows the same linear trend with ELM energy loss and the first ELM-resolved Ti measurements in the divertor region are shown. Measurements of flow shear and turbulence dynamics during L–H transitions show filaments erupting from the plasma edge whilst the full flow shear is still present. Off-axis neutral beam injection helps to strongly reduce the redistribution of fast-ions due to fishbone modes when compared to on-axis injection. Low-k ion-scale turbulence has been measured in L-mode and compared to global gyro-kinetic simulations. A statistical analysis of principal turbulence time scales shows them to be of comparable magnitude and reasonably correlated with turbulence decorrelation time. Te inside the island of a neoclassical tearing mode allow the analysis of the island evolution without assuming specific models for the heat flux. Other results include the discrepancy of the current profile evolution during the current ramp-up with solutions of the poloidal field diffusion equation, studies of the anomalous Doppler resonance compressional Alfvén eigenmodes, disruption mitigation studies and modelling of the new divertor design for MAST Upgrade. The novel 3D electron Bernstein synthetic imaging shows promising first data sensitive to the edge current profile and flows.

Progress toward fully noninductive discharge operation in DIII-D using off-axis neutral beam injection

The initial experiments on off-axis neutral beam injection into high noninductive current fraction (fNI), high normalized pressure (βN) discharges in DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] have demonstrated changes in the plasma profiles that increase the limits to plasma pressure from ideal low-n instabilities. The current profile is broadened and the minimum value of the safety factor (qmin) can be maintained above 2 where the profile of the thermal component of the plasma pressure is found to be broader. The off-axis neutral beam injection results in a broadening of the fast-ion pressure profile. Confinement of the thermal component of the plasma is consistent with the IPB98(y,2) scaling, but global confinement with qmin>2 is below the ITER-89P scaling, apparently as a result of enhanced transport of fast ions. A 0-D model is used to examine the parameter space for fNI=1 operation and project the requirements for high performance steady-state discharges. Fully noninductive solutions are found with 4<βN<5 and bootstrap current fraction near 0.5 for a weak shear safety factor profile. A 1-D model is used to show that a fNI=1 discharge at the top of this range of βN that is predicted stable to n=1, 2, and 3 ideal MHD instabilities is accessible through further broadening of the current and pressure profiles with off-axis neutral beam injection and electron cyclotron current drive.

Integrated magnetic and kinetic control of advanced tokamak plasmas on DIII-D based on data-driven models

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, , are described.

Power requirements for electron cyclotron current drive and ion cyclotron resonance heating for sawtooth control in ITER

13 MW of electron cyclotron current drive (ECCD) power deposited inside the q = 1 surface is likely to reduce the sawtooth period in ITER baseline scenario below the level empirically predicted to trigger neoclassical tearing modes (NTMs). However, since the ECCD control scheme is solely predicated upon changing the local magnetic shear, it is prudent to plan to use a complementary scheme which directly decreases the potential energy of the kink mode in order to reduce the sawtooth period. In the event that the natural sawtooth period is longer than expected, due to enhanced α particle stabilization for instance, this ancillary sawtooth control can be provided from >10MW of ion cyclotron resonance heating (ICRH) power with a resonance just inside the q = 1 surface. Both ECCD and ICRH control schemes would benefit greatly from active feedback of the deposition with respect to the rational surface. If the q = 1 surface can be maintained closer to the magnetic axis, the efficacy of ECCD and ICRH schemes significantly increases, the negative effect on the fusion gain is reduced, and off-axis negative-ion neutral beam injection (NNBI) can also be considered for sawtooth control. Consequently, schemes to reduce the q = 1 radius are highly desirable, such as early heating to delay the current penetration and, of course, active sawtooth destabilization to mediate small frequent sawteeth and retain a small q = 1 radius. Finally, there remains a residual risk that the ECCD + ICRH control actuators cannot keep the sawtooth period below the threshold for triggering NTMs (since this is derived only from empirical scaling and the control modelling has numerous caveats). If this is the case, a secondary control scheme of sawtooth stabilization via ECCD + ICRH + NNBI, interspersed with deliberate triggering of a crash through auxiliary power reduction and simultaneous pre-emptive NTM control by off-axis ECCD has been considered, permitting long transient periods with high fusion gain. The power requirements for the necessary degree of sawtooth control using either destabilization or stabilization schemes are expected to be within the specification of anticipated ICRH and ECRH heating in ITER, provided the requisite power can be dedicated to sawtooth control.

Sensitivity of transport and stability to the current profile in steady-state scenario plasmas in DIII-D

Recent experiments on DIII-D have provided the first systematic data on the impact of the current profile on the transport and stability properties of high-performance, steady-state scenario plasmas. In a future tokamak, to achieve 100% noninductive conditions and produce net power, the current profile J must be sustained by a large fraction of bootstrap current JBS, which is nonlinearly coupled with the kinetic profiles. Systematic scans of qmin and q95 were performed to determine empirically the best alignment of the noninductive currents with J and the variation of the transport properties with q. Transport analysis indicates that χe and χi are sensitive to the details of J in a way that makes the pressure profile peaking and JBS scale nonlinearly with both q and β in the experiment. Drift wave stability analysis yields linear growth rates that do not reproduce experimental trends in χ with qmin and q95. At high beta, necessary to maximize fBS, the plasma duration is often limited by n=1 tearing modes, whose stability also depends on the J profile. Broadly deposited electron cyclotron (EC) current at mid-radius was found to supply part of the required noninductive current and to positively affect the tearing stability. The modes appear when JEC is turned off for stable cases and always appear when the EC deposition is shifted outwards. The variation in the EC scan results is consistent with PEST3 calculations, showing that the tearing stability becomes extremely sensitive to small perturbations of the equilibrium in wall-stabilized plasmas run close to the ideal MHD limit. These modeling results are being used to design new experiments with higher ideal and tearing limits. A new capability for off-axis neutral beam injection system will be used to explore higher qmin scenarios and different current alignments.

Alfvén eigenmode structure during off-axis neutral beam injection

The spatial structure of Alfvén eigenmodes on the DIII-D tokamak is compared for contrasting fast ion deposition profiles resulting from on- and off-axis neutral beam injection (NBI). In both cases, poloidal mode rotation and eigenmode twist, or radial phase variation, are correlated with the direction of the normal ion diamagnetic flow and readily inverted with a reversal of toroidal magnetic field, BT. While off-axis NBI results in weakly driven reversed shear induced Alfvén eigenmodes due to reduced fast ion pressure gradient, ∇βfast, in the region of the mode, these marginally unstable modes exhibit a 2D phase structure that is indistinguishable from that observed during on-axis injection. This result is consistent with recent explorations using the non-perturbative codes Gyro and TAEFL that show a weak dependence of eigenmode structure on drive when fast ion density is uniformly reduced by a scalar multiplier. These codes also obtain unstable, counter-propagating modes with the inverted 2D phase structure when BT is kept constant and the diamagnetic flow direction is reversed by making ∇βfast sufficiently positive for an isotropic population of fast ions. While measurements of the spatial profile of fast ion D-α light from the recently upgraded charge exchange recombination diagnostic on DIII-D suggest a strong modification of fast ion pressure towards this limit, no counter-propagating modes have yet been observed in experiment.

Simulation of EAST off-axis neutral beam heating and current drive

The capability of off-axis neutral beam heating and current drive has been investigated with NUBEAM for Experimental Advanced Superconducting Tokamak (EAST). Three different approaches to realize off-axis Neutral Beam Injection (NBI) have been studied. Simulation results for on- and off-axis NBI are reported. The effects of the alignment of NBI relative to the magnetic field pitch on off-axis neutral beam heating and current drive are observed and discussed qualitatively. By comparing the numerical results, a most favorable off-axis NBI configuration is recommended. The capability to control sawtooth is also investigated by comparing locations of the q=1 rational surface and the peak of the fast ion density profile.

Fast-ion D-alpha measurements at ASDEX Upgrade

A fast-ion D-alpha (FIDA) diagnostic has been developed for the fully tungsten coated ASDEX Upgrade (AUG) tokamak using 25 toroidally viewing lines of sight and featuring a temporal resolution of 10 ms. The diagnostic's toroidal geometry determines a well-defined region in velocity space which significantly overlaps with the typical fast-ion distribution in AUG plasmas. Background subtraction without beam modulation is possible because relevant parts of the FIDA spectra are free from impurity line contamination. Thus, the temporal evolution of the confined fast-ion distribution function can be monitored continuously. FIDA profiles during on- and off-axis neutral beam injection (NBI) heating are presented which show changes in the radial fast-ion distribution with the different NBI geometries. Good agreement has been obtained between measured and simulated FIDA radial profiles in MHD-quiescent plasmas using fast-ion distribution functions provided by TRANSP. In addition, a large fast-ion redistribution with a drop of about 50% in the central fast-ion population has been observed in the presence of a q = 2 sawtooth-like crash, demonstrating the capabilities of the diagnostic.