Toroidal Rotation Profile Research Articles

Investigation on the roles of equilibrium toroidal rotation during edge-localized mode mitigated by resonant magnetic perturbations

The effects of equilibrium toroidal rotation during edge-localized mode (ELM) mitigated by resonant magnetic perturbation (RMP) are studied with the experimental equilibria of the EAST tokamak based on the four-field model in the BOUT++ code. As the two main parameters to determine the toroidal rotation profiles, the rotation shear and magnitudes were separately scanned to investigate their roles in the impact of RMPs on peeling–ballooning (P-B) modes. On one hand, the results show that strong toroidal rotation shear is favorable for the enhancement of the self-generated shearing rate with RMPs, leading to significant ELM mitigation with RMP in the stronger toroidal rotation shear region. On the other hand, toroidal rotation magnitudes may affect ELM mitigation by changing the penetration of the RMPs, more precisely the resonant components. RMPs can lead to a reduction in the pedestal energy loss by enhancing the multimode coupling in the turbulence transport phase. The shielding effects on RMPs increase with the toroidal rotation magnitude, leading to the enhancement of the multimode coupling with RMPs to be significantly weakened. Hence, the reduction in pedestal energy loss by RMPs decreased with the rotation magnitude. In brief, the results show that toroidal rotation plays a dual role in ELM mitigation with RMP by changing the shielding effects of plasma by rotation magnitude and affecting by rotation shear. In the high toroidal rotation region, toroidal rotation shear is usually strong and hence plays a dominant role in the influence of RMP on P-B modes, whereas in the low rotation region, toroidal rotation shear is weak and has negligible impact on P-B modes, and the rotation magnitude plays a dominant role in the influence of RMPs on the P-B modes by changing the field penetration. Therefore, the dual role of toroidal rotation leads to stronger ELM mitigation with RMP, which may be achieved both in the low toroidal rotation region and the relatively high rotation region that has strong rotational shear.

Study of correlations between LOC/SOC transition, intrinsic toroidal rotation reversal and TEM/ITG bifurcation with different working gases in TCV

The effects of different working gases on the transition from linear ohmic confinement (LOC) regime to saturated ohmic confinement (SOC) regime and its relation to the intrinsic toroidal rotation reversal phenomenon were explored in the TCV tokamak. The energy confinement saturation was studied across D, H and He density ramps, and a range of ECRH injection power and through variations of ohmic plasma current. The occurrence of rotation reversal, concomitantly with the LOC–SOC transition, was observed only for certain cases, making us formally exclude a causal relation between the two phenomena. A strong correlation between the evolution of toroidal rotation profiles and electron density gradients was, however, observed, in agreement with previous works (Lebschy et al 2017 Nucl. Fusion 58 026013; Hornsby et al 2018 Nucl. Fusion 58 056008). Linear gyrokinetic simulations were performed to probe the turbulent regime of these discharges, showing a dominance of trapped electron mode (TEM) during the LOC phase and a mixture of TEM and ion temperature gradient (ITG) following the transition to SOC regime in D. Such a TEM/ITG bifurcation was less pronounced in H and He. MHD activity was monitored throughout the discharges and possible correlations between sawteeth instability activity, energy confinement time saturation and rotation reversal are highlighted.

Fast-ion transport and toroidal rotation response to externally applied magnetic perturbations at the ASDEX Upgrade tokamak

This paper studies the effect of 3D magnetic perturbations (MPs) on fast-ion confinement, and its impact on the toroidal rotation velocity profile. Two low collisionality H-mode experiments carried out at the ASDEX Upgrade tokamak have been analysed. The two discharges feature different magnetic field helicity (q 95), and differences in the velocity-space and level of fast-ion losses are observed. A new analysis technique has been developed that sheds light on the dependencies between fast-ion losses and toroidal rotation, providing for the first time correlation patterns resolved in radius and velocity space of the lost fast-ions. The correlation intensifies towards the plasma edge and is strongly dependent on the orbit topology of the lost fast-ions. The ASCOT orbit following code has been used to characterize the fast-ion resonant transport and beam driven torques, using the vacuum approach and including plasma response (PR). The change of the toroidal canonical momentum, which serves as figure of merit for resonant fast-ion transport, has been calculated with ASCOT. The beam geometry and q 95 are found to have a strong impact on the fast-ion transport and losses. The fast-ion transport induced by the MPs affects the beam driven torques. The effect of the changes of the j × B and collisional torques on plasma rotation is analysed using the torques simulated by ASCOT and simple momentum balance calculations. For the low q 95 = 3.8 discharge, which benefits from a resonant amplification, we find excellent agreement with the measured variation of the toroidal velocity. For the high q 95 = 5.5 discharge, the inclusion of the PR improves the comparison with experimental data with respect to the vacuum estimation, but still some differences with experiments are observed. This suggests that other non-resonant effects could play a role for the determination of the toroidal rotation profile.

Physics studies of ADITYA & ADITYA-U tokamak plasmas using spectroscopic diagnostics

Several spectroscopic diagnostics encompassing the spectral emission range from the x-ray to near infrared (NIR) have been developed, installed and operated for diagnosing and physics studies in the ADITYA and ADITYA-U tokamaks. The recycling and impurity influxes and plasma Z eff after lithium (Li) coating have been studied using a PMT (photomultiplier tube)-filter based system by capturing H α , O1+, C2+, and visible continuum emissions. Significant reduction in the Z eff values has been observed in the discharge with the Li coated walls. The measured radial profile of H α emission using a filter-PMT array, has been modelled using a neutral transport code. The results show substantial contributions from the molecular hydrogen and molecular hydrogen ion dissociation (∼56%) and charge-exchange (∼30%) processes in the measured H α emission. Furthermore, a high-resolution, 1 m spectrometer with charge coupled device detector capable of multi-track measurements, has been used to study impurity transport, neutral and ion temperature and intrinsic plasma rotation. By modelling the measured radial profile of O4+ spectral line emission using an impurity transport code, substantial contribution of edge fluctuations on the oxygen transport has been observed. The toroidal ( ∼ 20 km s−1 in core) and poloidal ( ∼ 4.5 km s−1 at edge) rotation velocities are measured using C5+ (529 nm) and C2+ (464.7 nm) passive line emissions respectively. The measurement of radial profile of toroidal plasma rotation revealed a reversal of rotation direction depending on the electron density content of the ADITYA-U plasmas. The neutral temperature measurements showed a poloidal asymmetry indicating a presence of asymmetrical source of neutral heating. Moreover, the modelling of measured Fe14+ and Fe15+ vacuum ultraviolet spectral lines has revealed the neo-classical nature of iron transport in ADITYA core. Fast visible camera images captured the formation of filament structures triggered by interchange instabilities during plasma disruptions.

Influence of toroidal rotation on plasma response to external RMP fields in tokamak

The type-I edge localized mode (ELM) is a critical event associated with magneto-hydrodynamic(MHD) instabilities occurring in tokamak high-confinement (H-mode) discharges, that leads to huge heat loads on the plasma phasing components (PFC) and may result in material damages. It is important to effectively control large ELMs, in order to ensure safe operation of the future reactor-scale devices such as ITER and DEMO. Resonant magnetic perturbation (RMP) has been experimentally demonstrated to be a mature and robust technique for controlling ELMs. A set of parameters, such as the edge safety factor, the plasma flow, the RMP coil geometry and the spectrum of the applied external field, have been found to play important roles in controlling ELMs by RMP. Furthermore, the plasma pressure is known to affect the plasma response to the RMP field, in particular near the no-wall beta limit. This is because high plasma pressure drives the resonant field amplification of the external field by the plasma response. The ITER 10 MA steady state scenario will be operated near the no-wall stability limit. The new tokamak device HL-2M will also operate in the relatively high-beta regimes. On the other hand, more investigations are still needed to understand the influence of toroidal flow on the high-beta plasma response. This work employs a single fluid toroidal model to compute the plasma RMP response in HL-2M, emphasizing on the roles of two key physical quantities: the plasma resistivity and the toroidal rotation. The former allows penetration of the external RMP field into the plasma, while the latter mainly provides screening effect on the resonant field component. More specifically, the MARS-F code is utilized to study the plasma response to the externally applied <i>n</i> =1 ( n is the toroidal mode number) RMP field for high-beta HL-2M discharges, while varying the plasma toroidal rotation profile. The plasma response is found to (i) substantially modify the poloidal spectrum of the applied vacuum RMP field, (ii) change the amplitude of the resonant radial field amplitude near the plasma edge, and (iii) affect optimal current phasing between the two rows of RMP coils on HL-2M. A sufficiently slow toroidal flow near the plasma edge amplifies the radial field at rational surfaces associated with the perturbation. Since the latter serves as a reliable indicator for controlling the type-I edge localized mode (Type-I ELM) by RMP, varying rotation profile near the plasma edge offers a promising approach to optimize ELM control.

Impurity toroidal rotation profile measurement using upgraded high-resolution visible spectroscopic diagnostic on ADITYA-U tokamak.

A high-resolution spectroscopic diagnostic for the measurement of spatial profiles of impurity ion toroidal rotation velocities on the ADITYA-U tokamak has been upgraded to cover the complete plasma minor radius. Earlier, the coverage of diagnostics toward the plasma edge was restricted due to the placement of collection optics on the tangential port outside the vacuum vessel. The coverage of the full plasma minor radius, from 0 to 24cm, has been achieved using the newly designed and developed collection optics that have seven lines of sight to view the tokamak plasma mounted inside a customized re-entrant view port which is installed in the shadow of the limiter inside the vacuum vessel. The upgraded diagnostic also includes a faster charged coupled device detector with a smaller pixel size for the detection of a small wavelength shift. The complete spatial profile has been measured using the Doppler shifted passive change exchange spectral line at 529.0 nm from the C5+ ion. In this article, we present the collection optics' design, installation, calibration, and results obtained using the upgraded diagnostic.

Magnetic shear effect on plasma transport at T e/T i ∼ 1 through electron cyclotron heating in DIII-D plasmas

The effect of magnetic shear on plasma transport for an electron to ion temperature ratio (T e/T i) near unity has been explored in DIII-D utilizing electron cyclotron heating (ECH). Previous reports showed that significant confinement degradation occurred at T e/T i ∼ 1 in positive shear (PS) plasmas in DIII-D, whereas reduced confinement degradation was observed in negative central shear (NCS) plasmas. In this study, plasma transport in weak magnetic shear (WS) plasmas with ECH is investigated and compared with that in NCS and PS plasmas. Here the magnetic shears () are > 0.5, ∼0 and <-0.1 in the core region (ρ∼ 0.3–0.4) of PS, WS and NCS plasmas, respectively, and flat or negative inside ρ∼ 0.4 in the WS and NCS plasmas. Weak magnetic shear is found to be effective in minimizing degradation of ion thermal confinement as T e/T i increases through ECH application, and an improved confinement factor of H 98y2 ∼ 1.2 is maintained, similar to NCS plasmas. At T e/T i ∼ 1, the ion thermal diffusivity around an internal transport barrier decreases when changing the magnetic shear from positive to weak or negative shear. Also, reduced local particle and momentum transport was indicated by steeper density and toroidal rotation profiles in the weak and negative shear regimes. Linear gyrokinetic simulations predict little change in growth rates of low-k turbulence with ECH application in the WS and NCS plasmas, which is consistent with the transport and profile analyses.

Modeling plasma toroidal flow profile control via NTV torque with n = 2 3D fields in MAST-U

Toroidal modeling utilizing the MARS-Q code (Liu et al 2013 Phys. Plasmas 20 042503) shows that large (compared to electromagnetic torque) neoclassical toroidal viscous (NTV) torque can be achieved in the plasma core region in reference MAST(-U) L-mode plasmas, by applying the n = 2 (n is the toroidal mode number) 3D fields generated by the magnetic coils used for controlling the edge localized mode and/or by the error field correction coils. Large NTV torque, occurring at relatively slow plasma flow, in combination with strong variation of the torque amplitude versus the coil phasing of the 3D coils, offers a tool to control the plasma toroidal rotation profile in spherical tokamaks. To effectively employ the toroidal NTV torque, the potentially fast initial flow needs to first be damped by 3D fields. MARS-Q quasi-linear initial value simulations demonstrate that this is achievable in MAST-U, within the designed 3D coil current capability for the device.

Evidence and modeling of turbulence bifurcation in L-mode confinement transitions on Alcator C-Mod

Analysis and modeling of rotation reversal hysteresis experiments show that a single turbulent bifurcation is responsible for the Linear to Saturated Ohmic Confinement (LOC/SOC) transition and concomitant intrinsic rotation reversal on Alcator C-Mod. Plasmas on either side of the reversal exhibit different toroidal rotation profiles and therefore different turbulence characteristics despite the profiles of density and temperature, which are indistinguishable within measurement uncertainty. Elements of this bifurcation are also shown to persist for auxiliary heated L-modes. The deactivation of subdominant (in the linear growth rate and contribution to heat transport) ion temperature gradient and trapped electron mode instabilities is identified as the only possible change in turbulence within a reduced quasilinear transport model across the reversal, which is consistent with the measured profiles and inferred heat and particle fluxes. Experimental constraints on a possible change from strong to weak turbulence, outside the description of the quasilinear model, are also discussed. These results indicate an explanation for the LOC/SOC transition that provides a mechanism for the hysteresis through the dynamics of subdominant modes and changes in their relative populations and does not involve a change in the most linearly unstable ion-scale drift-wave instability.

Integrated Robust Control of the Global Toroidal Rotation and Total Plasma Energy in Tokamaks

Integrated-control solutions will play a significant role in future tokamaks, in which a variety of coupled control problems will need to be solved simultaneously by means of a limited number of actuators. In this article, the problem of simultaneously regulating the global toroidal rotation, Ω <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">φ</sub> , and total plasma energy, W, is tackled. These two 0-D variables, Ω <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">φ</sub> and W, depend on the ion toroidal rotation and electron temperature profiles, respectively. Both Ω <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">φ</sub> and W also depend on the electron density and safety factor profiles. The actuation methods considered in this article are co-current and counter-current neutral beam injection. A nonlinear, robust controller that makes use of Lyapunov redesign techniques is synthesized based on 0-D, control-oriented models of the Ω <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">φ</sub> and W dynamics. In addition, an actuator management scheme is designed to handle variations in the control priorities and availability of the neutral beam injectors. The actuator manager solves an optimization problem in real time in order to find the most appropriate course of action when unexpected changes occur. The integrated control architecture is tested for a DIII-D scenario by means of the 1-D code Control-Oriented Transport Simulator (COTSIM), which predicts the time evolution of the electron temperature, electron density, ion toroidal rotation, and safety factor profiles.

Predicting the rotation profile in ITER

Determining the toroidal rotation for future tokamaks like ITER is a challenging and important problem. By combining empirical scalings for the intrinsic rotation at the top of the pedestal with the expected neutral beam torque and modeling of momentum transport, the toroidal rotation profile for ITER is predicted with TGYRO using TGLF (SAT0 and SAT1). On axis rotation exceeds 20 krad s−1 and the shear is significant enough to reduce turbulent transport and significantly increase confinement and fusion power when comparing to cases that ignore the effect of rotation. The prediction of the rotation at the top of the pedestal is made with increased confidence due to experiments and modeling in DIII-D that have determined the importance of fast-ion and neutral particle transport effects on intrinsic rotation at this location. In particular, the effect of neutral particles on momentum transport in the pedestal region is found to be insignificant.

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.

Observations of toroidal plasma rotation reversal in the Aditya-U tokamak

An impurity ion toroidal rotation profile has been observed in the ohmic discharges of the Aditya-U tokamak from the Doppler-shifted passive charge exchange line emission of C5+ at 529 nm. The line has been monitored using a space-resolved visible spectroscopy diagnostic consisting of a 1 m multi-track spectrometer coupled with a charge coupled device (CCD) detector. In typical discharges with central chord-averaged electron density ≲2.5 × 1019 m−3, the maximum toroidal rotation velocity at the plasma core is found to be −20 km s−1 in the counter-current direction. Reversal of plasma toroidal rotation is observed for high electron density discharges with a central chord-averaged density of ~5 × 1019 m−3. During the rotation reversal, the direction of toroidal rotation changes to the co-current direction in the core. Interestingly, in the discharges with a central chord-averaged electron density of ≲2.5 × 1019 m−3, where neon gas is injected at the plasma current flat-top, the toroidal rotation reversal has been observed after the gas puff.

Hysteresis as a probe of turbulent bifurcation in intrinsic rotation reversals on Alcator C-Mod

Analysis and modeling of a new set of rotation reversal hysteresis experiments unambiguously show that changes in turbulence are responsible for the intrinsic rotation reversal and the linear to saturated ohmic confinement (LOC/SOC) transition on Alcator C-Mod. Plasmas on either side of the reversal exhibit different toroidal rotation profiles and therefore different turbulence characteristics despite profiles of density and temperature that are indistinguishable within measurement uncertainty. The deactivation of subdominant (in linear growth rate and heat transport) ion-temperature gradient and trapped electron mode-like instabilities in a mixed-mode state is identified as the only possible change in turbulence within a quasilinear transport approximation across the reversal which is consistent with the measured profiles and the inferred heat and particle fluxes. This indicates an explanation for the LOC/SOC transition that provides a mechanism for hysteresis through the dynamics of subdominant modes and changes in their relative populations, and does not involve a change in most (linearly) unstable ion-scale drift-wave instability.

Modeling 3D plasma boundary corrugation and tailoring toroidal torque profiles with resonant magnetic perturbation fields in ITER

Plasma response to 3D resonant magnetic perturbations (RMPs), applied for the purpose of controlling type-I edge localized modes (ELMs) in ITER with the baseline ELM control coils, is computed using a toroidal, resistive, full magneto-hydrodynamic model. Considered are five representative ITER plasmas, designed for different phases of the ITER exploration. The plasma response, measured by the plasma boundary corrugation, is found to be similar for the two DT scenarios at full plasma current (15 MA) and full toroidal field (5.3 T) but different fusion gain factors (Q = 5 versus Q = 10), indicating similar ELM control performance with the same RMP coil current configuration. The other plasma scenarios, with proportionally scaled down plasma current and toroidal field, can have different plasma boundary corrugation. The key plasma parameter affecting the response is the plasma toroidal flow near the pedestal region, which significantly varies depending on the transport model assumption for the toroidal momentum. Lower pedestal flow leads to a stronger edge peeling response from the plasma and thus probably a better ELM control. The optimal coil configuration for controlling type-I ELMs is similar for all four ITER plasmas with similar safety factor but different current levels, but is significantly different for the case at half plasma current (7.5 MA) and full field (5.3 T). On the other hand, for the purpose of controlling the radial profile of the plasma toroidal rotation in ITER using 3D fields, the relative amplitude of the toroidal torque density, between the plasma core and edge region, is optimized. Generally, a strong coupling between the core and edge torques is observed, largely due to the middle row ELM control coils. The best decoupling scheme of the core-edge torque distribution thus de-emphasizes the role of the middle row coils. Optimal coil current configurations are found for the ITER 15 MA/5.3 T Q = 10 plasma, that synergistically maximize the plasma edge-peeling response (indication for good ELM control) and the toroidal torque near the plasma edge (good for RMP field penetration through pedestal).

Understanding helium transport: experimental and theoretical investigations of low-Z impurity transport at ASDEX Upgrade

The presence of helium is fundamentally connected to the performance of a fusion reactor, as fusion-produced helium is expected to heat the plasma bulk, while He ‘ash’ accumulation dilutes the fusion fuel. An understanding of helium transport via experimentally validated theoretical models of the low-Z impurity turbulent transport is indispensable to predict the helium density profile in future fusion devices. At ASDEX Upgrade, detailed, multi-species investigations of low-Z impurity transport have been undertaken in dedicated experiments, resulting in an extensive database of helium and boron density profiles over a wide range of parameters relevant for turbulent transport (normalised gradients of the electron density, the ion temperature, and the toroidal rotation profiles, the collisionality and the electron to ion temperature ratio). Helium is not found to accumulate in the parameter space investigated, as the shape of the helium density profile follows largely that of the electron density. Helium is observed to be as peaked as the electron density at high electron cyclotron resonance heating fraction, and less peaked than the electron density at high neutral beam heating fraction. The boron density profile is found to be consistently less peaked than the electron density profile. Detailed comparisons of the experimental density gradients of both impurities with quasilinear gyrokinetic simulations have shown that a qualitative agreement between experiment and theory cannot always be obtained, with strong discrepancies observed in some cases.

Flux-driven integrated modelling of main ion pressure and trace tungsten transport in ASDEX Upgrade

Neoclassical and turbulent heavy impurity transport in tokamak core plasmas are determined by main ion temperature, density and toroidal rotation profiles. Thus, in order to understand and prevent experimental behaviour of W accumulation, flux-driven integrated modelling of main ion heat and particle transport over multiple confinement times is a vital prerequisite. For the first time, the quasilinear gyrokinetic code QuaLiKiz is applied for successful predictions of core kinetic profiles in an ASDEX Upgrade H-mode discharge in the turbulence dominated region within the integrated modelling suite JETTO. Neoclassical contributions are calculated by NCLASS; auxiliary heat and particle deposition profiles due to NBI and ECRH are prescribed from previous analysis with TRANSP. Turbulent and neoclassical contributions are insufficient in explaining main ion heat and particle transport inside the q = 1 surface, necessitating the prescription of further transport coefficients to mimic the impact of MHD activity on central transport. The ion to electron temperature ratio at the simulation boundary at stabilizes ion scale modes while destabilizing ETG modes when significantly exceeding unity. Careful analysis of experimental measurements using Gaussian process regression techniques is carried out to explore reasonable uncertainties. In following trace W impurity transport simulations performed with additionally NEO, neoclassical transport under consideration of poloidal asymmetries alone is found to be insufficient to establish hollow central W density profiles. Reproduction of these conditions measured experimentally is found possible only when assuming the direct impact of a saturated MHD mode on heavy impurity transport.

Main ion and impurity edge profile evolution across the L- to H-mode transition on DIII-D

Detailed measurements of the main ion () and impurity ion () evolution during the development of the H-mode pedestal across an L–H transition show significant differences in toroidal rotation, density, and temperature profiles in the pedestal region on DIII-D. While both species experience a slow toroidal spin up at constant input neutral beam injected torque, the toroidal rotation develops a non monotonic notch feature and lower toroidal rotation near the plasma edge immediately following the L–H transition. This feature is not present in the main ion rotation that instead, depending on plasma parameters, can show a flat or peaked rotation near the separatrix. The and temperature profiles show a similar evolution; however, the D+ temperature is lower than the temperature at the separatrix in both L and H-mode which may be due to cooling of via charge exchange with cold edge deuterium neutrals. Local neoclassical predictions of the main ion toroidal rotation based on the impurity properties show good agreement with direct measurements at the pedestal top for a lower power, higher collisionality case but can diverge significantly in the steep gradient region for the two shots studied here. These observations highlight the importance of directly measuring the properties of the main ion species at the plasma edge.

Tractable flux-driven temperature, density, and rotation profile evolution with the quasilinear gyrokinetic transport model QuaLiKiz

Quasilinear turbulent transport models are a successful tool for prediction of core tokamak plasma profiles in many regimes. Their success hinges on the reproduction of local nonlinear gyrokinetic fluxes. We focus on significant progress in the quasilinear gyrokinetic transport model QuaLiKiz (Bourdelle et al 2016 Plasma Phys. Control. Fusion 58 014036), which employs an approximated solution of the mode structures to significantly speed up computation time compared to full linear gyrokinetic solvers. Optimisation of the dispersion relation solution algorithm within integrated modelling applications leads to flux calculations faster than local nonlinear simulations. This allows tractable simulation of flux-driven dynamic profile evolution including all transport channels: ion and electron heat, main particles, impurities, and momentum. Furthermore, QuaLiKiz now includes the impact of rotation and temperature anisotropy induced poloidal asymmetry on heavy impurity transport, important for W-transport applications. Application within the JETTO integrated modelling code results in 1 s of JET plasma simulation within 10 h using 10 CPUs. Simultaneous predictions of core density, temperature, and toroidal rotation profiles for both JET hybrid and baseline experiments are presented, covering both ion and electron turbulence scales. The simulations are successfully compared to measured profiles, with agreement mostly in the 5%–25% range according to standard figures of merit. QuaLiKiz is now open source and available at www.qualikiz.com.

Deuterium charge exchange recombination spectroscopy from the top of the pedestal to the scrape off layer in H-mode plasmas

Recent completion of the thirty two channel main-ion (deuterium) charge exchange recombination spectroscopy (CER) diagnostic on DIII-D [J.L. Luxon, Nucl. Fusion 42 (2002) 614] enables detailed comparisons between impurity and main-ion temperature, density, and toroidal rotation. Sixteen sightlines cover the core of the plasma and another sixteen are densely packed towards the edge, providing high resolution measurements of the pedestal and steep gradient edge region of H-mode plasmas. The complexities of the Dα spectrum require fitting with a comprehensive model, as well as using iterative collisional radiative modeling to determine the underlying thermal deuterium ion properties. Large differences in the structure and magnitude of impurity (C6+) and main-ion (D+) toroidal rotation profiles are seen in the H-mode pedestal. Additionally the D+ temperature can be half the value of the C6+ temperature at the separatrix and shows more of a pedestal structure. Typically only the impurity properties are measured and the main-ion properties are either assumed to be the same, or inferred using neoclassical models, which require validation in the steep gradient region. These measured differences have implications for transport model validation, intrinsic rotation studies, pedestal stability, and the boundary conditions for scrape off layer and plasma material interactions studies.