Electron Internal Transport Barrier Research Articles

On the parasitic absorption in FWCD experiments in JET ITB plasmas

Fast wave current drive (FWCD) experiments have been performed in JET plasmas with electron internal transport barriers produced with LHCD. Because of a large fraction of parasitic absorption, owing to weak single pass damping, the inductive nature of the plasma current and the interplay between the RF-driven current and the bootstrap current only small changes are seen in the central current profiles. The measured difference in the central current density for co- and counter-current drive is larger than the response expected from current diffusion calculations, but smaller than the driven currents, suggesting a faster current diffusion than that given by neo-classical resistivity. A large fraction of the power is absorbed by cyclotron damping on residual 3He ions while a significant fraction appears not to have been deposited in the plasma. The strong degradation of heating and current drive occurs simultaneously with strong increases in the Be II and C IV line intensities in the divertor. The degradation depends on the phasing of the antennas and increases with reduced single pass damping which is consistent with RF-power being lost by dissipation of rectified RF-sheath potentials at the antennas and walls. Asymmetries in direct electron heating, lost power and production of impurities, fast ions and gamma-rays are seen for co- and counter-current drive. These differences are consistent with the differences in the absorption on residual 3He ions owing to the RF-induced pinch. Effective direct electron heating, comparable to the indirect electron heating with H-minority heating, occurs for dipole phasing of the antennas without producing a significant fast ion pressure and with low impurity content in the divertor plasma.

High-bootstrap, noninductively sustained electron internal transport barriers in the Tokamak à Configuration Variable

Important ingredients of the advanced-tokamak route to fusion have been explored in depth in the Tokamak à Configuration Variable [F. Hofmann, J. B. Lister, M. Anton et al., Plasma Phys. Controlled Fusion 36, B277 (1994)] over the past two years. Using a uniquely powerful and flexible electron-cyclotron resonance heating (ECRH) system as the primary actuator, fully noninductive, steady-state electron internal transport barrier discharges have been generated with an electron-energy confinement time up to five times longer than in L mode, poloidal β up to 2.4, and bootstrap fraction up to 75%. Interpretative transport modeling confirms that the safety factor profile is nonmonotonic in these discharges. The formation of the barrier is a discrete event resulting in rapid and localized confinement improvement consistent with the time and location of magnetic-shear reversal. In steady state, however, the confinement quality appears to depend on the current gradient in a broader negative-shear region enclosed by the barrier, improving with increasing shear: in particular, the width and depth of the barrier can be controlled and finely tuned, along a magnetohydrodynamic-stable path, by manipulating the current profile with ECRH (six independently steerable 0.45 MW launchers). The crucial role of the current profile has been clearly demonstrated by applying small Ohmic current perturbations which dramatically alter the properties of the barrier, enhancing or reducing the confinement with negative and positive current, respectively, with negligible Ohmic heating. These results are in agreement with theoretical estimates: first-principle-based numerical simulations of microinstability dynamics and turbulence-driven transport predict a substantial suppression of turbulence and anomalous energy diffusivity near the location of the minimum in the safety factor.

Inductive Current Density Perturbations to Probe Electron Internal Transport Barriers in Tokamaks

Improved electron energy confinement in tokamak plasmas, related to internal transport barriers, has been linked to nonmonotonic current density profiles. This is difficult to prove experimentally since usually the current profiles evolve continuously and current injection generally requires significant input power. New experiments are presented, in which the inductive current is used to generate positive and negative current density perturbations in the plasma center, with negligible input power. These results demonstrate unambiguously for the first time that the electron confinement can be modified significantly solely by perturbing the current density profile.

Electron ITB Formation with Combination of NBI and ECH in LHD

The electron internal transport barrier (ITB) is formed with centrally focused electron cyclotron resonance heating superposed on plasmas heated by neutral beam injection in the Large Helical Device. The electron transport is investigated for the electron ITB plasmas observed in various magnetic axis positions of Rax = 3.6, 3.75, and 3.9 m, and it turns out that the core electron transport is reduced with suppression of the anomalous transport in all three magnetic axis positions. In the theoretical calculations, positive radial electric fields are generated in the improved transport region, implying that the electron ITB formation is correlated with the neoclassical electron root. At an outer-shifted configuration of Rax = 3.9 m, where the helical ripple is large, the thermal diffusivity is decreased with decreasing collisionality, suggesting the reduction of the ripple transport by the radial electric field. The temperature and density conditions for the ITB formation are consistent with the theoretical density dependence of the transition temperature to the neoclassical electron root from the ion root.

Internal Transport Barrier and Bifurcation Phenomena in Stellarators

Bifurcation phenomena with confinement improvement in toroidal plasmas are briefly reviewed. After a general discussion of the formation mechanisms of improved confinement modes, the electron internal transport barrier commonly observed in stellarators is chosen here as the main subject with focus on the following aspects: the bifurcation property, the scenario of bifurcation with confinement improvement, and the confinement features. Several works to investigate the relationship between rational surfaces and barrier formation are introduced as examples to show the relation between magnetic topology and the bifurcation property. The roles of stellarators, which have a wide variety of configurations, for understanding bifurcation phenomena are discussed.

Chapter 6: Transport Studies in the FTU

Transport studies are presented in this chapter. Global scaling studies have been performed using several transport codes. Ohmic plasmas are found to follow the ITER97 L-mode scaling. Transport coefficients are discussed for improved confinement scenarios achieved in the Frascati Tokamak Upgrade (FTU): the repetitive pellet enhanced plasma mode, showing neoclassical confinement with H-factors up to 1.6, and the electron internal transport barriers (ITBs) with large transport barriers and H-factors up to 1.3. Heat transport models have been tested using electron cyclotron resonance heating (ECRH), steady or modulated, as a probe. The electron temperature stiffness observed in the main bulk of steady FTU plasmas can be interpreted both with a critical gradient transport model and with a model based on the existence of canonical profiles. ECRH has also been used to benefit from the improved confinement generally associated with low or negative magnetic shear, and large electron temperatures have been achieved in these conditions. Profile resiliency is observed so that heat transport is not consistent with a constant thermal diffusivity. Experimental optimization is discussed together with the analysis of transport coefficients. Thorough discussions of impurity transport are given, both for intrinsic and injected (from laser blow-off) impurities. Code simulation and experimental data are compared for a series of FTU experiments focusing on the improved confinement modes (pellets and ITBs). A moderate inward pinch velocity is generally required to reproduce the data.

Chapter 2: Highlights of the Physics Studies in the FTU

The main physics results achieved in the recent years in the Frascati Tokamak Upgrade (FTU) are reviewed. The main focus of research has been the development of performance plasmas at high densities (up to 4 × 1020 m-3), high magnetic field (up to 8 T) and plasma current (up to 1.6 MA), that are therefore in a domain of relevance for burning physics experiments such as ITER. The main tools consist in the development of plasma conditioning techniques and the use of various electron heating and current drive systems. Improved confinement regimes have been developed, including (a) the production of steady electron internal transport barriers at high density and electron temperature (up to central electron temperature of 11 keV at a central density of 0.9 × 1020 m3), (b) the production of repetitive pellet enhanced plasma modes with deep pellet deposition leading to a substantial increase of the neutron yield (and a record FTU value of the fusion product niTiτE up to 0.8 × 1020 m-3 keV·s), and (c) the production of radiation improved modes at high magnetic field. Main results on the supporting physics program will also be given in the domain of plasma wave physics (lower hybrid current drive, electron cyclotron resonance frequency, ion Bernstein waves), heat and impurities transport, and magnetohydrodynamic studies.

Chapter 3: Internal Transport Barrier Studies in the FTU

Strong electron internal transport barriers (ITBs) are obtained in the Frascati Tokamak Upgrade (FTU) with the combined injection of lower hybrid (LH) (up to 1.9 MW) and electron cyclotron (EC) (up to 0.8 MW) radio-frequency waves. ITBs occur during either the current plateau or the ramp-up phase, both in full and partial current drive (CD) regimes, up to ne0 > 1.4 × 1020 m-3, relevant to ITER operation. Central electron temperatures Te0 > 8 keV, at ne0 [approximately equal to] 0.8 × 1020 m-3, are sustained for up to 36 confinement times. The ITB extends over a region where a slightly reversed magnetic shear is established by off-axis LHCD and can be even larger than r/a = 0.5. EC power is used either to benefit from this improved confinement by heating inside the ITB or to enhance the peripheral LH power deposition and CD with off-axis resonance. Collisional ion heating is also observed, but thermal equilibrium with the electrons is not attained since the electron-ion equipartition time is always 4 to 5 times longer than the energy confinement time. An extensive transport modeling of these discharges, performed by means of the ASTRA code, is also presented. During the ITB phase, the ion diffusivity is close to the neoclassical value while the electron shear-dependent Bohm-gyro-Bohm model accounts quite well for Te(r,t), The Ray Tracing Fokker-Planck model, used to describe the LHCD physics, appears satisfactory to analyze and interpret the experimental results. It turns out that the barrier radius is mainly influenced by the LHCD deposition. In particular, a wider barrier is obtained the lower qa is and the larger the plasma density is.

Characteristics of transport in electron internal transport barriers and in the vicinity of rational surfaces in the Large Helical Device

Characteristics of transport in electron internal transport barriers (ITB) and in the vicinity of a rational surface with a magnetic island are studied with transient transport analysis as well as with steady state transport analysis. Associated with the transition of the radial electric field from a small negative value (ion-root) to a large positive value (electron-root), an electron ITB appears in the Large Helical Device [M. Fujiwara et al., Nucl. Fusion 41, 1355 (2001)], when the heating power of the electron cyclotron heating exceeds a power threshold. Transport analysis shows that both the standard electron thermal diffusivity, χe, and the incremental electron thermal diffusivity, χeinc (the derivative of normalized heat flux to temperature gradient, equivalent to heat pulse χe), are reduced significantly (a factor 5–10) in the ITB. The χeinc is much lower than the χe by a factor of 3 just after the transition, while χeinc is comparable to or even higher than χe before the transition, which results in the improvement of electron transport with increasing power in the ITB, in contrast to its degradation outside the ITB. In other experiments without an ITB, a significant reduction (by one order of magnitude) of χeinc is observed at the O-point of the magnetic island produced near the plasma edge using error field coils. This observation gives significant insight into the mechanism of transport improvement near the rational surface and implies that the magnetic island serves as a poloidally asymmetric transport barrier. Therefore the radial heat flux near the rational surface is focused at the X-point region, and that may be the mechanism to induce an ITB near a rational surface.

Control of electron internal transport barriers in TCV

Current profile tailoring has been performed by application of electron cyclotron heating (ECH) and electron cyclotron current drive, leading to improved energy confinement in the plasma core of the TCV tokamak. The improved confinement is characterized by a substantial enhancement (H-factor) of the global electron energy confinement time relative to the prediction of the RLW scaling law (Rebut P H et al 1989 Proc. 12th Int. Conf. of Plasma Physics and Controlled Fusion Research (Nice, 1988) vol 2 (Vienna: IAEA) p 191), which predicts well Ohmic and standard ECH discharges on TCV. The improved confinement is attributed to a hollow current density profile producing a reversed shear profile creating an electron internal transport barrier. We relate the strength of the barrier to the depth of the hollow current density profile and the volume enclosed by the radial location of the peak current density. The (Tresset G et al 2002 Nucl. Fusion 42 520) criterion is used to evaluate the performance of the barrier relative to changes in the ECH parameters or the addition of Ohmic current, which aid in identifying the control parameters available for improving either the strength or volume of the barrier for enhanced performance. A figure of merit for the global scaling factor is used that scales the confinement enhancement as the product of the barrier volume and strength.

Formation conditions for electron internal transport barriers in JT-60U plasmas

The formation of electron internal transport barriers (ITBs) was studied using electron cyclotron (EC) heating in JT-60U positive shear (PS) and reversed shear (RS) plasmas with scan of neutral beam (NB) power. With no or low values of NB power and with a small radial electric field (Er) gradient, a strong, box-type electron ITB was formed in RS plasmas while a peaked profile with no strong electron ITBs was observed in PS plasmas within the available EC power. When the NB power and the Er gradient were increased, the electron transport in strong electron ITBs with EC heating in RS plasmas was not affected, while electron thermal diffusivity was reduced in conjunction with the reduction of ion thermal diffusivity, and strong electron and ion ITBs were formed in PS plasmas.

Cold pulse experiments in plasma with an electron internal transport barrier on LHD

Transient transport experiments are performed in LHD plasma with electron internal transport barrier (e-ITB). Evidence for a reduction of electron heat diffusivity inside the ITB is observed from cold and heat pulse propagations. The observed enhancement of the cold pulse peak is explained by the temperature dependent electron heat diffusivity. The heat diffusivity inside the ITB decreases with an increase in the electron temperature in LHD.

Comparison of electron internal transport barriers in the large helical device and JT-60U plasmas

Plasmas with an electron internal transport barrier (ITB), which is characterized by peaked electron temperature profiles, are obtained in the JT-60U tokamak and in the large helical device (LHD) when the electron cyclotron heating (ECH) is focused on the magnetic axis. The maximum values of , where R is the major radius and is the scale length of the electron temperature gradient, are similar for the LHD and JT-60U ITB plasmas. However, there is a clear jump of observed in LHD but not in JT-60U in the ECH power scan. This result is consistent with the fact that the trigger mechanism of the electron ITB is the fast transition of the radial electric field from a small negative Er to a large positive Er in LHD and a change of the magnetic shear from positive to negative is required for the formation of the electron ITB in JT-60U. There are also differences in the electron temperature profiles inside the ITB. The flattening of the electron temperature profile inside the strong ITB could be explained by the sharp increase of q values observed in JT-60U, while no flattening of the electron temperature profile is observed in LHD, where the central q values stay low.

Study of impurity transport in FTU ITB plasmas

Impurity transport during FTU electron internal transport barrier (ITB) discharges has been studied by means of a one-dimensional time dependent transport model to reproduce plasma line and continuum emissions. The impurity behaviour has been explored in two different experimental frames in which the formation and maintenance of an ITB is obtained in FTU. In the first scenario the lower hybrid and electron cyclotron resonance heating waves are launched during the current (Ip) flat-top phase, while in the second scenario the RF power is injected early during the Ip ramp-up phase.A diffusion coefficient enhanced inside the barrier, by a factor of 10, and an inward pinch velocity linearly increasing with the radius up to 3.5 m s−1, as found in the Ohmic case, are necessary to reproduce FTU plasma emission in the first scenario. A diffusion coefficient lowered inside the barrier (by a factor of 2) and about the same inward pinch velocity (v(a) = 5 m s−1 in this case) have to be assumed to interpret the impurity behaviour if the RF power is injected during the Ip ramp-up phase. These diffusion coefficient profiles are similar to the ion thermal diffusivity profiles predicted by the JETTO code: an improved electron thermal conductivity at the centre but degraded ion thermal conductivity is predicted for the flat-top ITB case, while improved electron thermal conductivity in the centre with a slightly improved ion diffusion inside the barrier results in the ramp-up ITB.

Electron internal transport barrier formation and dynamics in the plasma core of the TJ-II stellarator

The influence of magnetic topology on the formation of electron internal transport barriers (e-ITBs) has been studied experimentally in electron cyclotron heated plasmas in the stellarator TJ-II. e-ITB formation is characterized by an increase in core electron temperature and plasma potential. The positive radial electric field increases by a factor of 3 in the central plasma region when an e-ITB forms. The experiments reported demonstrate that the formation of an e-ITB depends on the magnetic configuration. Calculations of the modification of the rotational transform due to plasma current lead to the interpretation that the formation of an e-ITB can be triggered by positioning a low order rational surface close to the plasma core region. In configurations without any central low order rational, no barrier is formed for any accessible value of heating power. Different mechanisms associated with neoclassical/turbulent bifurcations and kinetic effects are put forward to explain the impact of magnetic topology on radial electric fields and confinement.

An overview of results from the TCV tokamak

The Tokamak à Configuration Variable (TCV) tokamak (R = 0.88 m, a < 0.25 m, B < 1.54 T) programme is based on flexible plasma shaping and heating for studies of confinement, transport, control and power exhaust. Recent advances in fully sustained off-axis electron cyclotron current drive (ECCD) scenarios have allowed the creation of plasmas with high bootstrap fraction, steady-state reversed central shear and an electron internal transport barrier. High elongation plasmas, κ = 2.5, are produced at low normalized current using far off-axis electron cyclotron heating and ECCD to broaden the current profile. Third harmonic heating is used to heat the plasma centre where the second harmonic is in cut-off. Both second and third harmonic heating are used to heat H-mode plasmas, at the edge and centre, respectively. The ELM frequency is decreased by the additional power. In separate experiments, the ELM frequency can be affected by locking to an external perturbation current in the internal coils of TCV. Spatially resolved current profiles are measured at the inner and outer divertor targets by Langmuir probe arrays during ELMs. The strong, reasonably balanced currents are thought to be thermoelectric in origin.

Chapter 4: Advanced Tokamak Studies in ASDEX Upgrade

Advanced scenarios in tokamaks seek to maximize the confinement and stability of thermonuclear plasmas. Key to obtaining these conditions is operation at different current density profiles. Experiments at ASDEX Upgrade are reported with approximately zero magnetic shear in the center or reversed magnetic shear in the center. With zero magnetic shear and q0 near 1, stationary conditions are obtained in discharges without sawteeth at 800 kA and 1 MA and q95 = 3.3 to 4.5, using a combination of central neutral beam injection (NBI) heating and off-axis NBI heating. In this regime, the temperature profiles are stiff. Central heating with ion cyclotron resonance heating and electron cyclotron resonance heating can be used to prevent excessive density peaking to maximize the stability against neoclassical tearing modes and to prevent impurity accumulation. At a lower plasma current of 400 kA with 10 MW of NBI heating, the bootstrap current fraction in this regime is above 50% giving, with the NBI current drive, nearly fully noninductively driven conditions. Operation at average electron densities of 80 to 90% of the Greenwald density limit is obtained at a triangularity of δ = 0.43 achieving βN = 3.5 in stationary conditions. Moreover, in these plasmas, type II edge-localized modes are observed in configurations close to double null. In plasmas with a reversed magnetic shear in the center, the formation of ion transport barriers with NBI heating was optimized to obtain more reproducible transport barriers with an H-mode edge for maximum stability, achieving, transiently, βN values of 4. With a 1.6 MW counter electron cyclotron current drive in the center and densities in the range <ne> = 1.3 to 2.0 × 1019 m-3, a reversed magnetic shear and electron internal transport barriers are formed and sustained at 600 kA for 1 to 2 s with Te0 > 20 keV. Of the scenarios presented, the stationary plasmas with low magnetic shear in the center and q95 in the range 3.3 to 4.5 would obtain reactor-relevant values for H × βN/q952, a figure of merit used as a benchmark.

Modelling of the electron cyclotron current drive experiments in the TCV tokamak

With the very high electron cyclotron (EC) wave power density achieved in the TCV tokamak, more than 20 MW m−3, quasilinear modelling predicts an electron cyclotron current drive (ECCD) efficiency well in excess to the experimental value, by up to a factor of 10. Experimentally, radial transport of suprathermal electrons consistent with a diffusion coefficient larger than 1.5 m2 s−1 has been observed. This implies that the radial transport time is of the same order as the electron deflection time, suggesting that the key to resolving the discrepancy is to include radial transport in the kinetic simulations. In this paper we show that with a diffusion coefficient in accordance with the experimental estimation, we can reproduce the observed current drive efficiency in the fully EC current driven plasmas of TCV by solving the Fokker–Planck equation. Experimentally the total wave-driven current is well-known since the current in the Ohmic transformer is set to a constant value. We study the radial profile and the velocity dependence of the radial diffusion coefficient.A specific model is employed for steady-state electron internal transport barriers, produced by off-axis ECCD, with the electrons divided into two groups according to their energy. A small diffusion coefficient is assigned for the low-energy electrons, while at higher energies the diffusion level is chosen such as to obtain the experimental ECCD efficiency. The total current density, which is the sum of the wave-driven part and the bootstrap current, is found to be hollow, supporting the hypothesis that the reversed shear is the cause of the transport barrier.

Characteristics of electron heat transport of plasma with an electron internal-transport barrier in the large helical device.

Associated with the transition from ion root to electron root, an electron internal transport barrier (ITB) appears in the large helical device, when the heating power of electron cyclotron resonance heating exceeds the threshold power. The incremental thermal diffusivity of electron heat transport chi(inc)(e) in the ITB plasma is much lower than that in the plasma with the heating power below the threshold, and the thermal diffusivity chi(e) decreases with increasing of heating power [dchi(e)/d(P/n(e))<0] in helical ITB plasmas.

Progress towards internal transport barriers at high plasma density sustained by pure electron heating and current drive in the FTU tokamak

Strong electron internal transport barriers (ITBs) are obtained in FTU by the combined injection of lower hybrid (LH, up to 1.9 MW) and electron cyclotron (EC, up to 0.8 MW) radio frequency waves. ITBs occur during either the current plateau or the ramp-up phase, and both in full and partial current drive (CD) regimes, up to peak densities ne0>1.2×1020 m−3, relevant to ITER operation. Central electron temperatures Te0>11 keV, at ne0≈0.8×1020 m−3 are sustained longer than 35 confinement times. The ITB extends over a region where a slightly reversed magnetic shear is established by off-axis LHCD and can be as wide as r/a = 0.5. The EC power, instead, is used either to benefit from this improved confinement by heating inside the ITB, or to enhance the peripheral LH power deposition and CD with off-axis resonance. Collisional ion heating is also observed, but thermal equilibrium with the electrons cannot be attained since the e−–i+ equipartition time is always 4–5 times longer than the energy confinement time. The transport analysis performed with both ASTRA and JETTO codes shows a very good relation between the foot of the barrier and the weak/reversed shear region, which in turn depends on the LH deposition profile. The Bohm-gyroBohm model accounts for the electron transport until Te0<6 keV, but is pessimistic at higher temperatures, where often also a reduction in the ion thermal conductivity is observed, provided any magnetohydrodynamic activity is suppressed.