Impurity Density Profiles Research Articles

Calibration of the toroidal Charge eXchange Recombination Spectroscopy system on EAST

The toroidal Charge eXchange Recombination Spectroscopy (CXRS) diagnostic was recently deployed on EAST to measure plasma temperature, velocity and impurity density profiles. For spatial calibration, backlit fiber images were projected to a plane that intersected with the neutral beam centerline to accurately identify the measuring positions and angles. Wavelength calibration was performed using a neon lamp, which was also used to measure dispersions and instrumental functions of the spectrometers. To exclude the effect of fluctuating wavelength caused by ambient temperature, one fiber was fed by a standard lamp during the experiments to provide real-time wavelength calibration, which was complemented by the intrinsic impurity line emissions. Intensity calibration was done using an integrating sphere lamp. It was shown that the uncertainty of the position calibration was less than 10mm, while the angle and the wavelength was <0.5° and <0.005nm, respectively. These calibration results were applied in the data analysis to provide the plasma temperature, velocity and impurity density. Preliminary central carbon temperature was compared to argon temperature measured by the X-ray crystal spectrometer and agreements were observed.

Impurity confinement and transport in high confinement regimes without edge localized modes on DIII-Da)

Impurity transport in the DIII-D tokamak [J. L. Luxon, Nucl. Fusion 42, 614 (2002)] is investigated in stationary high confinement (H-mode) regimes without edge localized modes (ELMs). In plasmas maintained by resonant magnetic perturbation (RMP), ELM-suppression, and QH-mode, the confinement time of fluorine (Z = 9) is equivalent to that in ELMing discharges with 40 Hz ELMs. For selected discharges with impurity injection, the impurity particle confinement time compared to the energy confinement time is in the range of τp/τe≈2−3. In QH-mode operation, the impurity confinement time is shown to be smaller for intense, coherent magnetic, and density fluctuations of the edge harmonic oscillation than weaker fluctuations. Transport coefficients are derived from the time evolution of the impurity density profile and compared to neoclassical and turbulent transport models NEO and TGLF. Neoclassical transport of fluorine is found to be small compared to the experimental values. In the ELMing and RMP ELM-suppress...

Core impurity transport in Alcator C-Mod L-, I- and H-mode plasmas

Core impurity transport has been investigated for a variety of confinement regimes in Alcator C-Mod plasmas from x-ray emission following injection of medium and high Z materials. In ohmic L-mode discharges, impurity transport is anomalous (Deff ≫ Dnc) and changes very little across the LOC/SOC boundary. In ion cyclotron range of frequencies (ICRF) heated L-mode plasmas, the core impurity confinement time decreases with increasing ICRF input power (and subsequent increasing electron temperature) and increases with plasma current. Nearly identical impurity confinement characteristics are observed in I-mode plasmas. In enhanced Dα H-mode discharges the core impurity confinement times are much longer. There is a strong connection between core impurity confinement time and the edge density gradient across all confinement regimes studied here. Deduced central impurity density profiles in stationary plasmas are generally flat, in spite of large amplitude sawtooth oscillations, and there is little evidence of impurity convection inside of r/a = 0.3 when averaged over sawteeth.

Systematic Study of Tungsten Impurity Transport in Representative Regimes of Divertor Plasma

AbstractThe tungsten impurity transport has been studied by IMPGYRO code for a model tokamak geometry of JT‐60U tungsten experiments. The background plasma profiles calculated from SOLPS has been used. The density profiles of tungsten are compared between the low background deuterium density (Case A: n = 2.0 × 1019 m–3 at core side boundary) and the high background deuterium density (Case B: n = 3.0 × 1019 m–3 at core side boundary). In both cases, flow reversal in the SOL has been observed. The impurity density profile in the SOL is more localized for Case B. These features are possibly explained by the force balance between the thermal force and the friction force. (© 2014 WILEY‐VCH Verlag GmbH &amp; Co. KGaA, Weinheim)

Edge Radial Electric Field Formation after the L‐H Transition on JT‐60U

AbstractSpatio‐temporal measurements of the impurity ion temperature, density, and rotation profiles around the plasma edge region have been made in the JT‐60U tokamak, allowing the determination of radial electric field, Er, with the key dimensionless parameter (poloidal Mach number, Upm) at the L‐H transition in a number of operational regimes. We found that there is variation in the L‐H transition in terms of its time‐scale; not only “hard” type transition with a faster time‐scale than that seen in the plasma transport (as represented by an energy confinement time, τE) as seen in the many conventional tokamaks, but also “soft” one with a slow time‐scale (≈τE) is possible solution, including a complex multi‐stage Er transition in the later H‐phase. The most important point is that the critical condition for the L‐H transition predicted by ion‐orbit loss model could be applicable only for “hard” transition (occurred at Upm ≥ 1), and not necessary for “slow” one (occurred even at Upm &lt; 1). Characteristics of the turbulent density fluctuation with the frequency range of 100 kHz at the plasma edge region, in addition to a uniform toroidal MHD oscillation (i.e., n = 0), during ELM‐free H‐phase are also reported. (© 2014 WILEY‐VCH Verlag GmbH &amp; Co. KGaA, Weinheim)

Impurity transport studies in the Madison Symmetric Torus reversed-field pinch during standard and pulsed poloidal current drive regimes

The transport of intrinsic impurities is investigated during standard and improved confinement regimes of the Madison Symmetric Torus (MST) reversed-field pinch. The impurity diffusion coefficient (D) and pinch velocity (v) are obtained through comparing the time evolution of experimental impurity density profiles with the results of a one-dimensional impurity transport code. Experimental hollow fully stripped (C, O, B) ion populations in improved confinement discharges are reproduced with the transport code indicating outward convection of impurity ions. Estimated D and v are low and close to classical values. Standard MST discharges are characterized by a high level of stochasticity and nearly flat radial profiles of the fully stripped carbon. To reproduce this flat impurity profile a high outward convective velocity and high central D are assumed in the simulation.

Validation of x-ray line ratios for electron temperature determination in tokamak plasmas

X-ray imaging crystal spectroscopy (XICS) has been implemented on magnetic confinement fusion devices as a novel means of measuring local plasma temperature, impurity density, and flow profiles. At Alcator C-Mod, XICS allows for spatially-resolved, high spectral resolution measurements between 0.3 and 0.4 nm, enabling detailed analysis of He-like argon x-ray emission. Electron temperatures in the range of 0.5 keV ⩽Te ⩽3.0 keV are determined from He-like argon emissivity ratios of the n = 3 dielectronic satellites to the w-line and its surrounding n ⩾ 3 satellites, specifically the wavelength range of 3.9440 Å ⩽ λ ⩽ 3.9607 Å. These data are validated against measurements of Te from existing electron cyclotron emission and Thomson scattering diagnostics. Line ratio data are analysed via a tomographic inversion procedure, overcoming the traditional issue of spectra being averaged over the plasma cross-section. The implications of utilizing x-ray line ratios as a valid local temperature diagnostic are not limited to Alcator C-Mod; properties of plasma in future experiments as well as in astrophysical settings can also be investigated. The results of this experiment confirm that x-ray line ratios can be used as an accurate electron temperature diagnostic. The electron temperature can be determined from the relation to the line ratio, x, as Te[keV] = 0.1552x−0.7781 with 0.0223 < x < 0.2449.

Impurity behavior during sawtooth activity in tokamak plasmas

The transport of impurities by a sawtooth crash is simulated with the XTOR-2F code. Impurities are modeled as passive scalars, evolving in the compressible MHD flow inferred from the main MHD plasma. For a peaked impurity density profile, the non-linear kink flow of the sawtooth crash redistributes the profile efficiently and most of the particles in the peak inside the q = 1 surface are expelled. For an initially hollow impurity density profile, the crash leads to a significant penetration up to the magnetic axis. The results are compared with Kadomtsev's model. Despite essentially different mechanisms, the evolution of the particle content inside the q = 1 surface for Kadomtsev's model and for the non-linear case are virtually identical for the peaked profile, while the model slightly overestimates penetration for the hollow case.

Turbulent impurity transport modeling for Alcator C-Mod

AbstractTurbulent particle transport is investigated with a quasilinear theory that is motivated by the boron impurity transport experiments in the Alcator C-Mod. Eigenvalue problems for sets of reduced fluid equations for multi-component plasmas are solved for the self-consistent fluctuating field vectors composed of the electric potential φ, the main ion density δni, the impurity density δnz and the ion temperature fluctuation δTi. For Alcator C-Mod parameters, we investigate two drift wave models: (1) the density-gradient-driven impurity drift wave and (2) the ion-temperature-gradient-driven ion temperature gradient (ITG) mode. Analytic and numerical results for particle transport coefficients are derived and compared with the transport data and the neoclassical theory. We explore the ability of the model to explain impurity density profiles in three confinement regimes: H-mode, I-mode and internal transport barrier (ITB) regime in C-Mod. Related experiments reported on the Large Helical Device are briefly discussed.

Non-local heat transport, rotation reversals and up/down impurity density asymmetries in Alcator C-Mod ohmic L-mode plasmas

Several seemingly unrelated effects in Alcator C-Mod ohmic L-mode plasmas are shown to be closely connected: non-local heat transport, core toroidal rotation reversals, energy confinement saturation and up/down impurity density asymmetries. These phenomena all abruptly transform at a critical value of the collisionality. At low densities in the linear ohmic confinement regime, with collisionality ν* ⩽ 0.35 (evaluated inside of the q = 3/2 surface), heat transport exhibits non-local behaviour, core toroidal rotation is directed co-current, edge impurity density profiles are up/down symmetric and a turbulent feature in core density fluctuations with kθ up to 15 cm−1 (kθρs ∼ 1) is present. At high density/collisionality with saturated ohmic confinement, electron thermal transport is diffusive, core rotation is in the counter-current direction, edge impurity density profiles are up/down asymmetric and the high kθ turbulent feature is absent. The rotation reversal stagnation point (just inside of the q = 3/2 surface) coincides with the non-local electron temperature profile inversion radius. All of these observations suggest a possible unification in a model with trapped electron mode prevalence at low collisionality and ion temperature gradient mode domination at high collisionality.

High-resolution charge exchange measurements at ASDEX Upgrade

The charge exchange recombination spectroscopy (CXRS) diagnostics at ASDEX Upgrade (AUG) have been upgraded and extended to provide high-resolution measurements of impurity ion temperature, density, and rotation profiles. The existing core toroidal CXRS diagnostic has been refurbished to increase the level of signal, thus enabling shorter exposure times down to 3.5 ms. Additional lines of sight provide more detailed profiles and enable simultaneous measurements of multiple impurities. In addition, a new CXRS system has been installed, which allows for the measurement of poloidal impurity ion rotation in the plasma edge with high temporal (1.9 ms) and spatial resolution (down to 5 mm). A new wavelength correction method has been implemented to perform in situ wavelength calibrations on a shot-to-shot basis. Absolute measurements of the poloidal impurity ion rotation with uncertainties smaller than 1.5 km/s have been obtained. Comparison of all the CXRS measurements provides a consistency check of the diagnostics and good agreement has been found for all of the CXRS systems.

Method to obtain absolute impurity density profiles combining charge exchange and beam emission spectroscopy without absolute intensity calibration

Investigation of impurity transport properties in tokamak plasmas is essential and a diagnostic that can provide information on the impurity content is required. Combining charge exchange recombination spectroscopy (CXRS) and beam emission spectroscopy (BES), absolute radial profiles of impurity densities can be obtained from the CXRS and BES intensities, electron density and CXRS and BES emission rates, without requiring any absolute calibration of the spectra. The technique is demonstrated here with absolute impurity density radial profiles obtained in TEXTOR plasmas, using a high efficiency charge exchange spectrometer with high etendue, that measures the CXRS and BES spectra along the same lines-of-sight, offering an additional advantage for the determination of absolute impurity densities.

Model for Pedestal Transport Based on Suppression of Anomalous Transport Using ωE ×B Flow Shear and Magnetic Shear

A model for describing a pedestal transport in H -mode tokamak plasmas is developed and tested in BALDUR integrated predictive modelling code. The transport model is calculated based on the suppression of an anomalous core transport due to the ω E × B flow shear and magnetic shear. Because of the reduction of transport in the edge region, the pedestal can be formed and evolved. In this work, an anomalous transport is computed using a semi-empirical mixed Bohm/gyro-Bohm anomalous core transport model. The BALDUR code with both core and pedestal transport is used to simulate the time evolution of plasma current, ion and electron temperature, and particle and impurity density profiles for 10 DIII-D H -mode discharges in various plasma scenarios (i.e., gyro-radius scan, density scan, power scan, and elongation scan). It is found that the formation of the pedestal can be observed in all simulations, in which the L – H transition in the simulations is consistent with the L – H transition power threshold model. ...

Classical Impurity Ion Confinement in a Toroidal Magnetized Fusion Plasma

High-resolution measurements of impurity ion dynamics provide first-time evidence of classical ion confinement in a toroidal, magnetically confined plasma. The density profile evolution of fully stripped carbon is measured in MST reversed-field pinch plasmas with reduced magnetic turbulence to assess Coulomb-collisional transport without the neoclassical enhancement from particle drift effects. The impurity density profile evolves to a hollow shape, consistent with the temperature screening mechanism of classical transport. Corroborating methane pellet injection experiments expose the sensitivity of the impurity particle confinement time to the residual magnetic fluctuation amplitude.

Effect of poloidal asymmetry on the impurity density profile in tokamak plasmas

The effect of poloidal asymmetry of impurities on impurity transport driven by electrostatic turbulence in tokamak plasmas is analyzed. It is found that if the density of the impurity ions is poloidally asymmetric then the zero-flux impurity density gradient is significantly reduced and even a sign change in the impurity flux may occur if the asymmetry is sufficiently large. This effect is most effective in low shear plasmas with the impurity density peaking on the inboard side and may be a contributing factor to the observed outward convection of impurities in the presence of radio frequency heating.

Rotation and transport in Alcator C-Mod ITB plasmas

Internal transport barriers (ITBs) are seen under a number of conditions in Alcator C-Mod plasmas. Most typically, radio frequency power in the ion cyclotron range of frequencies (ICRFs) is injected with the second harmonic of the resonant frequency for minority hydrogen ions positioned off-axis at r/a > 0.5 to initiate the ITBs. They can also arise spontaneously in ohmic H-mode plasmas. These ITBs typically persist tens of energy confinement times until the plasma terminates in radiative collapse or a disruption occurs. All C-Mod core barriers exhibit strongly peaked density and pressure profiles, static or peaking temperature profiles, peaking impurity density profiles and thermal transport coefficients that approach neoclassical values in the core. The strongly co-current intrinsic central plasma rotation that is observed following the H-mode transition has a profile that is peaked in the centre of the plasma and decreases towards the edge if the ICRF power deposition is in the plasma centre. When the ICRF resonance is placed off-axis, the rotation develops a well in the core region. The central rotation continues to decrease as long as the central density peaks when an ITB develops. This rotation profile is flat in the centre (0 < r/a < 0.4) but rises steeply in the region where the foot in the ITB density profile is observed (0.5 < r/a < 0.7). A correspondingly strong E × B shear is seen at the location of the ITB foot that is sufficiently large to stabilize ion temperature gradient instabilities that dominate transport in C-Mod high density plasmas.

Radial electric field in JET advanced tokamak scenarios with toroidal field ripple

A dedicated campaign has been run on JET to study the effect of toroidal field (TF) ripple on plasma performance. Radial electric field measurements from experiments on a series of plasmas with internal transport barriers (ITBs) and different levels of ripple amplitude are presented. They have been calculated from charge exchange measurements of impurity ion temperature, density and rotation velocity profiles, using the force balance equation. The ion temperature and the toroidal and poloidal rotation velocities are compared in plasmas with both reversed and optimized magnetic shear profiles. Poloidal rotation velocity (vθ) in the ITB region is measured to be of the order of a few tens of km s−1, significantly larger than the neoclassical predictions. Increasing levels of the TF ripple are found to decrease the ion temperature gradient in the ITB region, a measure for the quality of the ITB, and the maximum value of vθ is reduced. The poloidal rotation term dominates in the calculations of the total radial electric field (Er), with the largest gradient in Er measured in the radial region coinciding with the ITB.

Observations of counter-current toroidal rotation in Alcator C-Mod LHCD plasmas

Following the application of lower hybrid current drive (LHCD) power, the core toroidal rotation in Alcator C-Mod L- and H-mode plasmas is found to increment in the counter-current direction, in conjunction with a decrease in the plasma internal inductance, li. Along with the drops in li and the core rotation velocity, there is peaking of the electron and impurity density profiles, as well as of the ion and electron temperature profiles. The mechanism generating the counter-current rotation is unknown, but it is consistent in sign with an inward shift of energetic electron orbits, giving rise to a negative core radial electric field. The peaking in the density, toroidal rotation (in the counter-current direction) and temperature profiles occurs over a time scale similar to the current relaxation time but slow compared with the energy and momentum confinement times. Most of these discharges exhibit sawtooth oscillations throughout, with the inversion radius shifting inward during the LHCD and profile evolution. The magnitudes of the changes in the internal inductance and the central rotation velocity are strongly correlated and found to increase with increasing LHCD power and decreasing electron density. The maximum effect is obtained with a waveguide phasing of 60° (a launched parallel index of refraction n‖ ∼ 1.5), with a significantly smaller magnitude at 120° (n‖ ∼ 3.1), and with no effect for negative or heating (180°) phasing. Regardless of the plasma parameters and launched n‖ of the waves, there is a strong correlation between the rotation velocity and li changes, possibly providing a clue for the underlying mechanism.

Spatially resolved high resolution x-ray spectroscopy for magnetically confined fusion plasmas (invited)

The use of high resolution x-ray crystal spectrometers to diagnose fusion plasmas has been limited by the poor spatial localization associated with chord integrated measurements. Taking advantage of a new x-ray imaging spectrometer concept [M. Bitter et al., Rev. Sci. Instrum. 75, 3660 (2004)], and improvements in x-ray detector technology [Ch. Broennimann et al., J. Synchrotron Radiat. 13, 120 (2006)], a spatially resolving high resolution x-ray spectrometer has been built and installed on the Alcator C-Mod tokamak. This instrument utilizes a spherically bent quartz crystal and a set of two dimensional x-ray detectors arranged in the Johann configuration [H. H. Johann, Z. Phys. 69, 185 (1931)] to image the entire plasma cross section with a spatial resolution of about 1 cm. The spectrometer was designed to measure line emission from H-like and He-like argon in the wavelength range 3.7 and 4.0 A with a resolving power of approximately 10,000 at frame rates up to 200 Hz. Using spectral tomographic techniques [I. Condrea, Phys. Plasmas 11, 2427 (2004)] the line integrated spectra can be inverted to infer profiles of impurity emissivity, velocity, and temperature. From these quantities it is then possible to calculate impurity density and electron temperature profiles. An overview of the instrument, analysis techniques, and example profiles are presented.

Impurity Transport in Alcator C-Mod Plasmas

Trace nonrecycling impurities (scandium and CaF2) have been injected into Alcator C-Mod plasmas in order to determine impurity transport coefficient profiles in a number of operating regimes. Recycling Ar has also been injected to characterize steady-state impurity density profiles. Subsequent impurity emission has been observed with spatially scanning X-ray and vacuum ultraviolet spectrometer systems, in addition to very high spatial resolution X-ray and bolometer arrays viewing the plasma edge. Measured time-resolved brightness profiles of helium-, lithium-, and beryllium-like transitions have been compared with those calculated from a transport code that includes impurity diffusion and convection, in conjunction with an atomic physics package for individual line emission. Similar modeling has been performed for the edge observations, which are unresolved in energy. The line time histories and the profile shapes put large constraints on the impurity diffusion coefficient and convection velocity profiles. In L-mode plasmas, impurity confinement times are short (~20 ms), with diffusivities in the range of 0.5 m2/s, anomalously large compared to neoclassical values. During Enhanced Dα (EDA) H-modes, the impurity confinement times are longer than in L-mode plasmas, and the modeling suggests that there exists inward convection (≤50 m/s) near the plasma edge, with greatly reduced diffusion (of order 0.1 m2/s), also in the region of the edge transport barrier. These edge values of the transport coefficients during EDA H-mode are qualitatively similar to the neoclassical values. In edge localized mode-free H-mode discharges, impurity accumulation occurs, dominated by large inward impurity convection in the pedestal region. A scaling of the impurity confinement time with H-factor reveals a very strong exponential dependence. In internal transport barrier discharges, there is significant impurity accumulation inside of the barrier foot, typically at r/a = 0.5. Steady-state impurity density profiles in L-mode plasmas have a large up-down asymmetry near the last closed flux surface. The impurity density enhancement, in the direction opposite to the ion B × ΔB drift, is consistent with modeling of neoclassical parallel impurity transport.