DIII-D edge physics data base

Modeling of carbon transport in the divertor and SOL of DIII-D during high performance plasma operation

SOLPS modelling of ELMing H-mode

Scrape-off layer measurements in DIII-D

Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician's point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings.

Studies of processes in the scrape-off layer (SOL) of the tokamak show a direct influence of periphery effects on confinement parameters of the plasma core. This paper illustrates experimentally observed transport barrier formation initialized by the low hybrid (LH) heating of the hydrogen plasma. The experimental data near last close flax surface (LCFS) and in SOL were obtained by means of an enhanced movable multielectrode Langmuir probe and spatially resolved spectroscopic technique retooled with additional helium puffing. The edge diagnostics show a strong change of a radial electric field and plasma parameters near LCFS. The alteration at the plasma edge is generated by high LH ion heating. It was found that the L-H transition is accompanied by a noticeable reconstruction of the poloidal and radial plasma parameters profiles in the SOL and in the limiter shadow region.

Although the research on divertors and scrape-off layers (SOLs) has been not as focused on as the recent success of the Korean fusion program, a few linear plasma devices have been developed for simulating divertor and SOL plasmas: (1) diversified plasma simulator (DiPS), a versatile linear machine, has been developed for simulations of divertor and space plasmas with various electric probes, such as single, triple, and Mach Probes and gridded energy analyzer. DiPS consists of two major parts: a divertor plasma simulator with a \(\hbox {LaB}_6\) DC plasma source and a space plasma simulator with a helicon RF plasma source, (2) divertor plasma simulator-1 (DiPS-1) is a part of DiPS with only a \(\hbox {LaB}_6\) cathode, where a low-power laser-induced fluorescence (LIF) is added and more electric probe diagnostics are augmented; it is dedicated only for fusion edge and divertor plasmas, (3) Divertor Plasma Simulator-2 (DiPS-2) has been modified from the DiPS-1 by adding a magnetic nozzle with a limiter structure and by removing the helicon source and space chamber. DiPS-2 is a linear plasma device with a 4-inch \(\hbox {LaB}_6\) cathode, the same as DiPS-1, and it is focused on the development of various diagnostics, such as those used for LIF and laser Thomson scattering (LTS) along with various electric probes, on the divertor and scrape-off plasmas and on the plasma-material interaction (PMI) research, such as that of tungsten and graphite as plasma-facing components (PFCs), (4) A Multi-Purpose Plasma (\(\hbox {MP}^2\)) device is a renovation of the Hanbit mirror device [Kwon et al., Nucl. Fusion 43, 686 (2003)] with the installation of two plasma sources: \(\hbox {LaB}_6\) (DC) and helicon (RF) plasma sources. A honeycomb-like large-area \(\hbox {LaB}_6\) (HLA-\(\hbox {LaB}_6\)) cathode has been developed for the divertor plasma simulation to improve the resistance against the thermal shock fragility for large (8-inch) and high density plasma generation, (5) DiPS-2 has been augmented by adding another cylindrical device, called the Dust interaction with Surfaces Chamber (DiSC) for the generation and diagnostics of dusts. This combined system (DiPS-2+DiSC) has added two more diagnostics: Laser Photo-Detachment (LPD) for dust density and laser Mie Scattering (LMS) for dust size. Moreover, dusts or negative ions have been analyzed by using electric probes and capacitive diagram gauges in Transport and Removal of Dusts (TReD) device.

This work aims at modeling the ELMing H-mode of the experimental advanced superconducting Tokamak (EAST) using edge plasma code package SOLPS5.0. The steady-state H-mode is obtained by adjusting the perpendicular radial anomalous transport coefficient according to the given upstream profiles of experimental measurements (shot 36291). The downstream divertor region can be then modeled once the simulated upstream electron density and temperature agrees well with the experimental measurements. On the basis of the H-mode simulation, the effect of the drift on divertor targets power asymmetry and that of the power damping width of scrape-off layer (SOL) on the peak power density to the divertor target are modeled, the simulation results indicate that the drift (E×B, B×▽B) is one of the main reasons for targets power asymmetry, and the increasing of power damping width of the SOL will reduce the peak power load on the divertor target sharply; besides, due to the presence of interaction between plasma and neutral gas or radiation in divertor region, the effect of the power damping width on the divertor target power load is decreased.

To explore the influence of drifts on ion flows in the DIII-D divertor and Scrape-Off-Layer (SOL), a calibrated Coherence Imaging System (CIS) views the lower divertor. A new tuneable calibration laser and precision wavemeter provide the rest wavelength for CIII 465 nm plasma emissions resulting in ≤ 1 km/s uncertainty in ion velocities that are in the range of 30 km/s in the divertor plasma. The 2-D line-integrated velocity images have been tomographically inverted to obtain maps of ion velocity. Both He + and C 2 + ion velocity maps were measured in both directions of the toroidal magnetic field BT, and C 2 + images were obtained in attached and detached L- and H-mode plasmas. The measured C 2 + velocities are in the range of 20–30 km/s and are compared with UEDGE modeling with full drifts. While He II as a main ion is expected to be entrained in the plasma flow, the C 2 + impurity flow involves the competition between the frictional and ion temperature gradient forces. Both the CIS data and UEDGE modeling agree that the C 2 + flow is towards the divertor plates, but flips with the direction of BT. In general, the experimental C 2 + emission and flow profiles have a wider spatial extent than the UEDGE modeling.

In this article, recent research progress of laser induced fluorescence (LIF) diagnostics in ASIPP is reviewed. The fundamental goal of ASIPP's effort on the LIF diagnostics is to develop a working velocity distribution function (VDF) diagnostic to monitor helium ash removal in burning plasmas. Real-time monitoring of helium ash VDF requires consideration in radiation safety, measurement response time, and compatibility of tokamak operation, as well as ion/neutral VDFs with a temperature at possibly >20 eV. These considerations, almost unique to large-scale instruments or specifically to tokamak plasmas, require renewed research and development (R&D) efforts in optical design, automation, and expansion of laser functionality for the LIF diagnostics that were often not considered in conventional applications of the LIF diagnostics. To this end, ASIPP invested a series of research effort into the development of LIF techniques. Signal evaluation has been performed on a diagnostics-test device (LTS), and the obtained signal strength has been extrapolated to tokamak edge and divertor relevant environments with promising results. The extrapolations compared the available solid angle and plasma density between the test device and the design scenario where the LIF diagnostics will potentially be implemented in a tokamak edge and divertor plasma. The estimation neglects the higher density of energetic electrons in the edge and divertor regions, making it a conservative assessment. Automated post-DAQ processing of LIF signals using both “conventional” methods and an AI-augmented method were also explored. Both approaches yielded usable results, but the AI-trained method showed a faster response, which is advantageous for real-time plasma diagnostics. Basic mechanisms of LIF diagnostics were also explored. Specifically, the limitations of lock-in modulation was demonstrated to determine the limitation of time-resolved measurements using such methods. De-modulation and degradation of the LIF signal occurred as the modulation frequency exceeded 1/10th of the fluorescence frequency. This finding sets a practical limit of the modulation frequency up to which we can benefit from lock-in amplification. These published works along with other unpublished progress will be thoroughly discussed in this article to illustrate ASIPP's effort towards the implementation of LIF diagnostics in future tokamak devices and to promote the application of LIF diagnostics in other fields of research.

Comparison between an open divertor and a more-closed divertor in DIII-D demonstrates detachment up to 40% lower pedestal density (ne,ped) in the closed divertor due to a combination of decreased fueling of the pedestal and increased dissipation in the scrape off layer (SOL) in the closed divertor, both resulting from increased neutral trapping in the divertor. Predicting whether the relationship between divertor closure and detachment will hold for an opaque SOL, in which the contribution of ionizing neutrals to fueling the pedestal is lessened, requires separating out different mechanisms contributing to the density difference at detachment. A series of experiments on DIII-D characterizes matched discharges using various divertor configurations to isolate the effects of divertor closure. These experiments show detachment up to 25% lower ne,sep in the closed divertor than in the open divertor, supported by simulations showing increased neutral trapping, and hence, increased dissipation, in the closed divertor. A difference in ne,ped/ne,sep is also seen: for matched ne,sep, the closed divertor has up to 20% lower ne,ped, consistent with modeling showing a smaller ionization fraction inside the separatrix in this case. Understanding how these pieces fit together will help in the development of predictive models of pedestal density and detached divertors compatible with a high performance core.

Possible influence of near SOL plasma on the H-mode power threshold

Calculations of the profiles along the field lines within the divertor and scrape-off layer (SOL) of differences in the plasma ion density, temperature, parallel current, parallel flow velocity, and electrostatic potential, which result from using different poloidal distributions of the particle and heat influxes crossing the separatrix from the core plasma into the SOL, are presented and discussed vis-à-vis experimental observations. The calculations show that the stronger outboard than inboard particle and heat fluxes into the SOL caused by the geometric compression/expansion of flux surfaces predicted by magnetohydrodynamic equilibrium calculations lead to a prediction of higher plasma temperature at the outboard divertor than at the inboard divertor, a result that is consistent with experimental observation and that confirms a previous prediction (made without accounting for drifts) of a possible cause of the observed in-out divertor power asymmetry. The calculations also illustrate the effect of the poloidal distribution of particle and power influx into the SOL on the flow velocity, parallel current, and electrostatic potential distributions in the SOL and divertor.

The radial electric field in the plasma edge is studied in the Large Helical Device (LHD) experiments. When magnetic field lines become stochastic or open at the plasma edge and connected to the vessel, electrons are lost faster than ions along these field lines. Then, a positive electric field appears in the plasma edge. The radial electric field profile can be used to detect the effective plasma boundary. Magnetic topology is an important issue in stellarator and tokamak research because the 3D boundary has the important role of controlling MHD edge stability with respect to ELMs, and plasma detachment. Since the stochastic magnetic field layer can be controlled in the LHD by changing the preset vacuum magnetic axis, this device is a good platform to study the properties of the radial electric field that appear with the different stochastic layer width. Two magnetic configurations with different widths of the stochastic layer as simulated in vacuum are studied for low-β discharges. It has been found that a positive electric field appeared outside of the last closed flux surface. In fact the positions of the positive electric field are found in the boundary between of the stochastic layer and the scrape-off layer. To understand where is the boundary of the stochastic layer and the scrape-off layer, the magnetic field lines are analyzed statistically. The variance of the magnetic field lines in the stochastic layer is increased outwards for both configurations. However, the skewness, which means the asymmetry of the distribution of the magnetic field line, increases for only one configuration. If the skewness is large, the connection length becomes effectively short. Since that is consistent with the experimental observation, the radial electric field can be considered as an index of the magnetic topology.

It has long been recognized that the plasma edge and the interaction of the plasma with material interfaces through the scrape-off layer (SOL) is crucial for the successful operation of future fusion machines, e.g. ITER. Tremendous progress has been achieved in the last decade in diagnosing, modelling and understanding this plasma region. Nonetheless, many important questions remain unanswered, e.g. the origin of parallel SOL flows, of ELM types or even of the L–H transition itself. While scientists often specialize and work either on the confined edge region or on the open magnetic field line region (SOL) it is obvious that the unsolved problems in the field can only be addressed by treating the pedestal and SOL regions in an integrated manner, thereby capturing their dynamical interaction.The present cluster issue is intended to provide an actual overview of the field of plasma edge and scrape-off layer both from the simulation/theory and the experimental side. The papers originate from contributions made to the IEA Large Tokamak IA Workshop on `Edge Transport in Fusion Plasmas' which was held in Krakow, Poland, 11–13 September 2006, in the cultural centre `Dom Polonii' (http://www.etfp2006.ifpilm.waw.pl). The workshop was attended by 30 participants from all over the world, many of whom are leading experts in their respective areas of work. The presentations focused on both experimental and theoretical aspects, from edge turbulence, edge localized modes (ELMs), to challenges in edge modelling using fluid and kinetic codes. It was a goal of the meeting, which was held in a convivial atmosphere of scientific debate, to bring together the progress made in diagnostics with the current state of the art in theory and code development.The nine papers in this cluster issue will provide an actual snapshot of this exciting and quickly evolving field. They cover a large number of themes, from edge turbulence measurements (Zweben) and simulation (Scott), via experimental (Kamiya) and numerical (Pankin) investigations of ELMs, towards developments of integrated transport codes (Kawashima) and a discussion of challenges facing edge modelling in the future (Schneider). The contributions also deal with topical themes of SOL flows (Zagorski), magnetic field ergodization (Jakubowski) and liquid lithium limiters (Pericoli). It is hoped that they provide compact access and a point of reference for the rapidly progressing work towards scientific understanding and finally control of the plasma edge and SOL region.

Lower hybrid current drive (LHCD) is a potential source of non-inductive off-axis current drive (CD) for tokamaks. Although LHCD has been successfully deployed on a number of tokamaks, it is highly sensitive to the scrape-off layer (SOL) conditions local to the LHCD launcher. Large gaps between the launcher and plasma core, SOL turbulence, or edge density perturbations due to edge-localized modes can hamper CD or cause large reflected power. These coupling issues in part motivated the installation of an LHCD launcher on the high-field side (HFS) of DIII-D. On the HFS, the SOL is less turbulent and more controllable compared to the low-field side. This quiescence may result in more predictable edge conditions and thus a more predictable CD. In this work, HFS SOL reflectometry measurements are predicted from global plasma parameters using machine learning models. The SOL predictions coupled with the full-wave simulation of the LHCD launcher allow for the prediction of reflected power, directivity, and arcing risk before the discharge. Launcher performance is then optimized using multi-objective Bayesian optimization, finding the shot parameters that result in an optimal SOL density that maximizes CD while minimizing the risk of arcing. The predictions and optimizations of LHCD performance are then accelerated using a surrogate model of the full-wave LHCD simulation.