Internal Transport Barrier Discharges Research Articles

Geometric dependencies of the mean E × B shearing rate in negative triangularity tokamaks

This paper presents a comparative study of the poloidal distribution of the mean E × B shearing rate for positive triangularity (PT) and negative triangularity (NT) tokamaks. The effects of flux surface up–down asymmetry due to asymmetric upper and lower triangularities are also considered. Both direct eddy straining and the effects on Shafranov shift feedback loops are examined. Shafranov shift increases the shearing rate at all poloidal angles for all triangularities, due to flux surface compression. The maximum shearing rate bifurcates at a critical triangularity . Thus, the shearing rate is maximal off the outboard midplane for NT, while it is maximal on the outboard midplane for PT. For up–down asymmetric triangularity, the usual up–down symmetry of the shearing rate is broken. The shearing rate at the outboard midplane is lower for NT than for PT, suggesting that the shearing efficiency in NT is reduced. Implications for turbulence stabilisation and confinement improvement in high- NT and internal transport barrier discharges are discussed.

Impact of fast ions on microturbulence and zonal flow dynamics in HL-2A internal transport barriers

The turbulent transport properties and dynamics of zonal flows (ZFs) in the presence of fast ions (FIs) are investigated for a typical internal transport barrier (ITB) plasma based on the gyrokinetic approach, focusing on the role of FI temperature and the effects of the toroidal rotation, including the E× B rotational shear, parallel velocity gradient (PVG) as well as the rotation velocity itself. Linear GENE simulations have shown that the core ITB plasma on HL-2A is dominated by ion temperature gradient (ITG) modes and trapped electron modes (TEMs), where the former is stabilized by FIs whereas destabilized by the PVG. Neither of the FIs or the PVG has observable effect on TEMs. The ion heat transport generally decreases at large FI temperature due to the nonlinear electromagnetic stabilization of turbulence with increased total plasma β until electromagnetic modes are excited. The transport fluxes peak around a certain FI temperature and the ZF shearing rate is significantly higher at such value compared with that in the absence of FIs, and the heat flux reduction is a result of the synergistic interaction between turbulence, ZFs and the external rotational shear. The E× B shear stabilizing and PVG destabilizing is not obvious at low normalized ITG R/L Ti, indicating they are less important in determining the stiffness level in the relatively low density and rotation scenarios regarding the HL-2A ITB discharges. The turbulence suppression is predominated by the nonlinear stabilization of ITG turbulence as well as enhanced ZFs simultaneously in the presence of FIs. These results have also provided the possible way to reduce the turbulence transport through increasing the FI temperature in the off-axis neutral beam heated plasmas such as in HL-2A.

Investigation of the role of fishbone activity in the formation of internal transport barrier in HL-2A plasma

A tokamak scenario with q(0) close to 1 has been achieved on HL-2A tokamak, which is an internal transport barrier (ITB) at low central shear or a steady-state ITB combined with an H-mode edge barrier. In this scenario, the formation of an ITB with a steep ion temperature profile is observed to be closely linked to the q = 1 magnetic surface and magnetohydrodynamic (MHD) activities around it, such as long-lived mode (LLM) or fishbone activities. Experimental evidence and simulation analysis suggest that the fishbone activities can induce a poloidal flow, which is beneficial for the suppression of turbulence in the plasma core region. Furthermore, an application of central electron cyclotron resonance heating (ECRH) to such beam-heated weak shear ITB discharges leads to a substantial effect on central MHD stability, converting the LLM into fishbone activity and hence enhancing the strength of the ITB with a much stronger gradient. Moreover, ITBs in combination with an H-mode barrier were achieved for 10 confinement times with βN = 2.0.

Sustainable internal transport barrier discharge at KSTAR

We explore the inboard-limited internal transport barrier (ITB) as an alternative advanced operation scenario for KSTAR. This paper presents in detail the progress of the ITB experiment at KSTAR. In an earlier study, the ITB formed in both ion and electron thermal channels, and an early neutral-beam injection (NBI) power of over 4 to 5 MW under a limited L-mode was crucial to ITB access. In the present study, we access the ITB experimentally with about 3 MW of NBI power by using the upper single null (USN), which is an unfavorable H-mode condition with a higher L–H power threshold. Finding an ITB access condition with a lower heating power should allow for a more efficient control of the heat flux and for maintaining stable plasma performance. The key control parameters of the experiment are the vertical position and the location of outboard striking point of the plasma. The shape-control attempts to divert the plasma to a vertically shifted USN with a marginal touch of the inboard limiter so that the plasma can remain in the L-mode at the boundary, while the striking-point control maintains the ITB performance for a longer period of time.

KSTAR Status and Upgrade Plan Toward Fusion Reactor

Korea Superconducting Tokamak Advanced Research (KSTAR) program has made significant advances in the development of long pulse and high-performance plasma scenarios, utilizing the advantages of the fully superconducting tokamak. Ten years of KSTAR operation have produced outstanding progress in plasma control, extending the operating window of the plasma discharges, and achieving the H-mode up to 1.2 MA in plasma current, up to 90 s at top duration, and up to 2.16 in elongation. The unique features of the KSTAR device also produced a record 30 s long pulse in H-mode operation without an edge-localized-mode (ELM) crash, and progress in the fundamental transport physics through systematic study, using unique diagnostic capabilities. Therefore, the next near-term research focus will be on how to extend the high performance plasmas up to 300 s by overcoming scientific and engineering obstacles to steady state operation. For fusion reactors, the tokamak device needs to sustain a permanent plasma current. It is also known that the bootstrap current should play an important role in the plasma current for an efficient fusion reactor, and its contribution should be over as high as possible. An internal transport barrier discharge based on the reverse shear profile of the safety factor is one of the promising candidates for enhancing the bootstrap current. However, even in reverse shear mode, an additional external off axis current drive (CD) should be provided to control the current profile, as well as to supplement the plasma current. KSTAR has pursued steady state to operation, aiming for a fully non inductive state, based on real time (RT) plasma control by maximizing the long pulse capabilities of the superconducting machine. In order to deposit current at an arbitrary position, the first requirement is to have many highly efficient CD actuators capable of profile diversity. KSTAR has prepared several RF CD portfolios in addition to a conventional on and off axis CD neutral beam injection (NBI) system. The final target of an off axis current drive will be a 4 MW off-axis NB, 8 MW on-axis NB, 6 MW electron cyclotron heating (ECH), and 4 MW radio frequency heating and current drive (RFHCD) (ion cyclotron (IC), lower hybrid (LH), Helicon CD). In addition, as a long-term goal, demonstrating a long pulse core edge compatible plasma in a reactor relevant condition is important in present medium size tokamak, since it is expected to sustain a very high pressure pedestal height plasma using super H-mode, with parallel upgrades of the superconducting toroidal field (TF) coils.

Formation of the internal transport barrier in KSTAR

One of key objectives of tokamak experiments is the exploration of enhanced confinement regimes, and the access of the internal transport barrier (ITB) formation is dealt with an important physics issue in the most of major tokamaks. Also, the advanced tokamak scenario with ITB is expected to lead to a continuous reactor with high fusion power density. From that point of view, the formation of the ITB in KSTAR which is designed for long pulse operation capability is very important although its heating and current drive systems are not fully equipped yet. We have therefore assumed that an early injection of the full NBI power (∼5.5 MW) during the current ramp-up would give a chance to form an internal barrier if the plasma could stay in the L-mode. To avoid the H-mode transition, we have produced inboard limited plasmas with detaching from the both upper and lower divertors. Using this approach, an ITB formation during L-mode has been observed which shows improved core confinement. Ion and electron temperature profiles show the barrier clearly in the temperature, and it was sustained for about 7 s in the dedicated experiment. This is the first stationary ITB observed in a full superconducting tokamak. This operation scenario with the ITB could be an alternative way to achieve a high performance regime in KSTAR, and the length of the ITB discharge could be extended even longer. In this paper, we present the formation of the ITB using measured and simulated characteristic profiles.

Design of tangential viewing phase contrast imaging for turbulence measurements in JT-60SA.

A tangential viewing phase contrast imaging system is being designed for the JT-60SA tokamak to investigate microturbulence. In order to obtain localized information on the turbulence, a spatial-filtering technique is applied, based on magnetic shearing. The tangential viewing geometry enhances the radial localization. The probing laser beam is injected tangentially and traverses the entire plasma region including both low and high field sides. The spatial resolution for an Internal Transport Barrier discharge is estimated at 30%-70% of the minor radius at k = 5 cm-1, which is the typical expected wave number of ion scale turbulence such as ion temperature gradient/trapped electron mode.

Neutron spectroscopy measurements of tritium beam transport at JET

A detailed description of the 14 MeV neutron emission in plasmas heated by neutral beam injection has been carried out by coupling Monte Carlo calculations of the neutron emission spectrum with TRANSP modelling of the beam ion energy distributions. The model is used to study tritium beam injection experiments of the JET trace tritium campaign for internal transport barrier (ITB) and H-mode discharges. For ITB discharges, the measured neutron emission spectrum is well described by modelling using as input the beam ion distribution calculated with TRANSP. For H mode discharges the neutron spectrum can be reproduced only if high energy tritons are lost from the plasma, suggesting the possible role of low frequency tearing modes on the beam ions. The presented results are of relevance for tritium beam transport studies in trace tritium experiments and, more generally, for deuterium and tritium transport studies in high power experiments using neutron emission spectroscopy.

Measurement and physical interpretation of the mean motion of turbulent density patterns detected by the beam emission spectroscopy system on the mega amp spherical tokamak

The mean motion of turbulent patterns detected by a two-dimensional beam emission spectroscopy (BES) diagnostic on the mega amp spherical tokamak (MAST) is determined using a cross-correlation time delay method. Statistical reliability of the method is studied by means of synthetic data analysis. The experimental measurements on MAST indicate that the apparent mean poloidal motion of the turbulent density patterns in the lab frame arises because the longest correlation direction of the patterns (parallel to the local background magnetic fields) is not parallel to the direction of the fastest mean plasma flows (usually toroidal when strong neutral-beam injection is present). This effect is particularly pronounced in a spherical tokamak because of the relatively large mean rotation and large magnetic pitch angle. The experimental measurements are consistent with the mean motion of plasma being toroidal. The sum of all other contributions (mean poloidal plasma flow, phase velocity of the density patterns in the plasma frame, non-linear effects, etc) to the apparent mean poloidal velocity of the density patterns is found to be negligible. These results hold in all investigated L-mode, H-mode and internal transport barrier discharges. The one exception is a high-poloidal-beta (the ratio of the plasma pressure to the poloidal magnetic field energy density) discharge, where a large magnetic island exists. In this case BES detects very little motion. This effect is currently theoretically unexplained.

Current density evolution in electron internal transport barrier discharges in TCV

Simulations of the plasma current density evolution in electron internal transport barrier discharges on the Tokamak à Configuration Variable (TCV) have been performed, in order to determine the relationship between the safety-factor profile and the electron transport improvement. The results show that the formation of the transport barrier is correlated with the shear reversal in all cases studied, regardless of the different heating and current drive schemes. No indications were found of discrete effects related to low-order rational q surfaces. On the contrary, the increase in confinement along with the negative shear is gradual, but constant, indicating that the transition is smooth, although it can be very fast.

Comparison of fusion alpha performance in JET advanced scenario and H-mode plasmas

Currently, plasmas with internal transport barriers (ITBs) appear the most likely candidates for steady-state scenarios for future fusion reactors. In such plasmas, the broad hot and dense region in the plasma core leads to high fusion gain, while the cool edge protects the integrity of the first wall. Economically desirable large bootstrap current fraction and low inductive current drive may, however, lead to degraded fast ion confinement. In this work the confinement and heating profile of fusion alphas were compared between H-mode and ITB plasmas in realistic JET geometry. The work was carried out using the Monte Carlo-based guiding-center-following code ASCOT.For the same plasma current, the ITB discharges were found to produce four to eight times more fusion power than a comparable ELMy H-mode discharge. Unfortunately, also the alpha particle losses were larger (∼16%) compared with the H-mode discharge (7%). In the H-mode discharges, alpha power was deposited to the plasma symmetrically around the magnetic axis, whereas in the current-hole discharge, the power was spread out to a larger volume in the plasma center. This was due to wider particle orbits, and the magnetic structure allowing for a broader hot region in the centre.

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.

Fully predictive time-dependent transport simulations of ITB plasmas in JET, JT-60U and DIII-D

For the first time, the predictive capabilities of the mixed Bohm/GyroBohm, Weiland and ‘retuned’ GLF23 transport models are investigated with ITB discharges from the ITPA ITB database with fully predictive, time-dependent transport simulations. A range of plasma conditions is examined for JET, JT-60U and DIII-D discharges with internal transport barriers (ITBs). The simulations show that the Bohm/GyroBohm model is able to follow the time evolution of the discharge from the preheating phase without an ITB through the ITB onset phase until the high performance phase with fair accuracy in most cases in JET and JT-60U. This indicates the importance of the interplay between the magnetic shear and ωE×B flow shear in ITB formation since these are the mechanisms that govern the ITB physics in the model. In order to achieve good agreement in DIII-D, the α-stabilization had to be included in the model, emphasizing the role played by the α-stabilization in the physics of the ITBs. The Weiland and GLF23 transport models show limited agreement between the model predictions and experimental time evolution of the ITB and kinetic plasma profiles. The Weiland model does not form a clear ITB in any of the three tokamaks despite varying plasma profiles, such as the q-profile. On the other hand, the average temperatures and density are often in fair agreement with experimental values. The GLF23 model often predicts an ITB, but its radial location is often too far inside the plasma, or shrinks as the simulations proceed in time. Consequently, the central temperatures at the end of the simulations during the high performance phase are usually underestimated. It is worth noting that GLF23 features in general better predictions of the Te and Ti profiles outside the ITB than the other models studied. Achieving the quantitative capability to predict the multi-channel ITB dynamics with fully predictive, time-dependent transport simulations has turned out to be extremely challenging.

Timescale and magnitude of plasma thermal energy loss before and during disruptions in JET

In this paper we analyse and discuss the thermal energy loss dynamics before and during JET disruptions that occurred between 2002 and 2004 in discharges which reached >4.5 MJ of thermal energy. We observe the slow thermal energy transients with diamagnetic loops and the fast ones with electron cyclotron emission and soft x-ray diagnostics. For most disruption types in JET, the plasma thermal energy at the time of the thermal quench is substantially less than that of the full performance plasma, typically in the range of 10–50% depending on plasma conditions and disruption type. The exceptions to this observation are disruptions in plasmas with a strong internal transport barrier (ITB) and in discharges terminating in a pure vertical displacement event, in which the plasma conserves a very high energy content up to the thermal quench. These disruption types are very sudden, leaving little scope for the combined action of soft plasma landing strategies and intrinsic performance degradation, both requiring >500 ms to be effective, to decrease the available thermal energy. The characteristic time for the loss of energy from the main plasma towards the PFCs in the thermal quench of JET disruptions is in the range 0.05–3.0 ms. The shortest timescales are typical of disruptions caused by excessive pressure peaking in ITB discharges. The available thermal energy fraction and thermal quench duration observed in JET can be processed (with due caution) into estimates for the projected PFC lifetime of the ITER target.

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.

Predicting core and edge transport barriers in tokamaks using the GLF23 drift-wave transport model

The density and temperature profiles are predicted in core and edge transport barriers in the DIII-D tokamak [J. L. Luxon and L. G. Davis, Fusion Technol. 8, 441 (1985)] using the GLF23 drift-wave model. The GLF23 model has been retuned to yield a better fit to the linear gyrokinetic growth rates for reversed magnetic shear and H-mode pedestal parameters. The turbulent saturation levels are determined using nonlinear gyrokinetic simulations. Using a large profile database, it is found that the retuned and original GLF23 models yield comparable results for discharges with monotonic safety factor profiles and no discernable internal transport barriers (ITBs). Examples of using retuned GLF23 model to predict the temperature profiles in simulations of several DIII-D strongly reversed magnetic shear ITB discharges are provided. Particle transport simulations show that the model is successful in predicting the density profile in discharges without ITBs but that some additional background particle diffusivity is needed in order to reproduce the measured density profiles within the barrier region of ITB plasmas where the ion temperature gradient and trapped electron mode transport have been quenched by rotational shear stabilization.

Helium exhaust experiments on JET with Type I ELMs in H-mode and with Type III ELMs in ITB discharges

An analysis of helium exhaust experiments on JET in the MkII-GB divertor configuration is presented. Helium is pumped by applying an argon frost layer on the divertor cryo pump. Measurement of the helium retention time, , is performed in two ways: by the introduction of helium in gas puffs and measurement of the subsequent decay time constant of the helium content, ; and by helium beam injection and measurement of the helium replacement time, . In ELMy H-mode, with plasma configuration optimized for pumping, is achieved, where is the thermal energy replacement time. For quasi-steady internal transport barrier (ITB) discharges, the achieved is significantly lower. The achieved helium recycling coefficient, confirmed by an independent measurement to be Reff ≈ 0.91, is the same in both scenarios. None of the discharges are dominated by core confinement. The difference in is instead due to the confinement properties of the edge plasma, which is characterized by Type I ELMs for the H-mode discharges studied, and Type III ELMs for the quasi-steady ITB discharges. This difference is quantified by an independent measurement of the ratio of the helium replacement time with a helium edge source to the energy confinement time.

Tritium transport experiments on the JET tokamak

An overview is given of the experimental method, the analysis technique and the results for trace tritium experiments conducted on the JET tokamak in 2003. Observations associated with events such as sawtooth collapses, neo-classical tearing modes and edge localized modes are described. Tritium transport is seen to approach neo-classical levels in the plasma core at high density and low q95, and in the transport barrier region of internal transport barrier (ITB) discharges. Tritium transport remains well above neo-classical levels in all other cases. The correlation of the measured tritium diffusion coefficient and convection velocity for normalized minor radii r/a = [0.65, 0.80] with the controllable parameters q95 and plasma density are found to be consistent for all operational regimes (ELMy H-mode discharges with or without ion cyclotron frequency resonance heating, hybrid scenario and ITB discharges). Scaling with local physics parameters is best described by gyro-Bohm scaling with an additional inverse beta dependence.

Formation criteria and positioning of internal transport barriers in ASDEX Upgrade

Advanced tokamak discharges on ASDEX Upgrade can generate an internal transport barrier (ITB) during the current ramp-up phase early in the discharge. Formation of the ITB has become more reliable with the discovery that a low density is necessary for it to form. These ITBs form in very low or negative central shear regions. There is no clear evidence of integer q magnetic surfaces triggering the ITB phase. The appearance of an integer q surface (usually q = 2) often leads to a second ITB (i.e. another steepening in the ion temperature gradient) inside the first ITB.Investigation of possible turbulence suppression mechanisms suggests that something is missing from the generally accepted model of sufficient E × B sheared flow suppressing the turbulence since the E × B shearing rate was not greater than, or comparable with, the maximum linear growth rate in about half of the ITB discharges analysed. Thus, it does not explain the reduced radial transport observed in these discharges.

Transport analysis of ohmic, L-mode and improved confinement discharges in FTU

A thorough investigation of confinement in Frascati Tokamak Upgrade has been carried out on a new database of ohmic, L-mode and advanced scenario discharges (multiple pellet-fuelled, radiation improved and internal transport barriers (ITBs)) obtained with the available auxiliary heating systems, namely electron cyclotron resonant heating, lower hybrid and ion Bernstein wave. A general agreement of the measured τE with ITER97 L-mode scaling is found in ohmic and L-mode discharges. An improvement of the energy confinement time (τE) of up to about 60% over the ITER97 L-mode scaling has been obtained in ITB discharges, together with a reduction in local electron transport in the region of high pressure gradient, and up to about 30% in pellet-fuelled discharges (where τE as large as ∼120 ms have been reached). The linear density dependence of τE in ohmic discharges has been found to extend above the saturation density threshold in pellet-fuelled plasmas.