Edge Plasma Code Research Articles

Computational Assessment on Impurity Sourcing and Transport Using High-Temperature Graphite and Silicon Carbide Plasma-Facing Walls in a Tokamak Environment

Operating with hot tokamak plasma-facing components will be essential in fusion reactors to maximize the thermal efficiency of the blanket. The SOLPS-ITER edge plasma code package and the DIVIMP Monte Carlo impurity tracking code were used in tandem to simulate the effect of active wall heating on impurity sourcing and transport in a DIII-D–size tokamak. The SOLPS-ITER plasma background was generated based on a previous DIII-D discharge and includes the effect of particle drifts. DIVIMP simulations found that actively heating the lower divertor (versus the divertor shelf or the entire wall) was the most efficient way to minimize gross erosion and core impurity influx at temperatures above 1000 K. Replacing the graphite wall with a silicon carbide (SiC) wall yielded a 5 to 20× decrease in the estimated gross erosion rate of carbon, with a maximum decrease observed at a lower divertor temperature of 800 K. Gross erosion of Si from SiC was estimated to be almost 100× lower than that of C from SiC, due primarily to the low impact energy of incident D plasma on the divertor targets. The core impurity influx for SiC walls is predicted to be lower than that with graphite walls, but eroded Si ions appear to migrate preferentially (versus C) to the core due to a more peaked erosion profile closer to the strike points where the ion temperature gradient force drives particles upstream. These predictive simulations suggest that active heating of the plasma-facing wall may both lower wall erosion and improve core performance relative to the “warm” walls of current devices that are typically only heated via plasma contact. Relative reductions in gross erosion and upstream accumulation by using SiC instead of graphite as the wall material strengthen the argument for upgrades to current graphite-clad machines and continued development of SiC first-wall and blanket concepts.

Recent progress in modeling ICRF-edge plasma interactions with application to ASDEX Upgrade

This paper summarizes recent progress in modeling the interaction between ion cyclotron range of frequency (ICRF) waves and edge plasma with application to ASDEX Upgrade. The basic theories, the development of ICRF and edge plasma codes, the integrated modeling methods and some key results are reviewed. In particular, the following physical aspects are discussed: (1) ICRF power coupling; (2) slow wave propagation; (3) ICRF-rectified sheath; (4) ICRF-induced convection; (5) ICRF-edge turbulence interaction. Moreover, comprehensive integrated modeling strategies by including all necessary codes in one package and solving multiple physical issues self-consistently are discussed.

Numerical simulation of fueling pellet ablation and transport in the EAST H-mode discharge

To understand the effect of injected deuterium (D) pellets on background plasma, the ablation of D pellets and the transport of D species in both atomic and ionic states in the EAST device are simulated using a modified dynamic neutral gas shield model combined with the edge plasma code SOLPS-ITER. The simulation results show that there is a phenomenon of obvious atomic deposition in the scrape-off layer (SOL) after pellet injection, which depends strongly on the injection velocity. With increasing injection velocity, the atomic density in the SOL decreases evidently and the deposition time is relatively shortened. Possible effects for triggering of edge localized modes (ELMs) by D and Li pellets are also discussed. With the same pellet size and injection velocity, the maximum perturbation pressure caused by D pellets is obviously higher. It is found that the resulting maximum perturbed pressure is remarkably enhanced when the injection velocity is reduced from 300 m/s to 100 m/s for a pellet with a cross section of 1.6 mm, which indicates that the injection velocity is important for ELM pacing. This work can provide reasonable guidance for choosing pellet parameters for fueling and ELM triggering.

Design of EAST lower divertor by considering target erosion and tungsten ion transport during the external impurity seeding

To demonstrate the performance of tungsten (W) as the divertor target material and to solve the power handling problem during high power long-pulse discharge, the upgrade of EAST lower divertor is planned. In this work, the physical design of the W divertor is presented by using 2D edge plasma code SOLPS and Monte Carlo impurity transport code DIVIMP. The optimized divertor geometry is proposed after systematic examination of target shapes, target slant angles and the pump opening locations. The performance of the designed divertor is further assessed by impurity seeding. By comparing the medium and high power discharges with argon (Ar) seeding, the differences on the divertor power radiation and impurity core accumulation are distinguished. The simulated effective ion charge Z eff fits well the scaling law, which is based on multi-machine database. Ar seeding and neon (Ne) seeding scans are carried out separately. The simulation results indicate Ar has higher power radiation efficiency than that of Ne, thus promoting the achievement of plasma detachment. However, the core compatibility with Ar is worse than with Ne. The W target erosion and W impurity transport during impurity seeding are simulated by the DIVIMP–SOLPS coupled modeling. It illustrates that under the similar divertor plasma conditions, Ar seeding causes more serious W erosion and more severe core contamination by W impurity, than Ne seeding. Finally, the divertor in–out asymmetry is studied by considering electromagnetic drifts. The simulation results manifest that the designed open vertical inner target reduces in–out asymmetry due to that its weak power radiation capability is offset by the ion flow driven by the drifts. In addition, the designed divertor is compatible with the quasi snowflake magnetic configuration. These studies will improve the understanding of W target sputtering and W impurity transport control during the radiative divertor discharges for CFETR/DEMO.

SOLPS analysis of changes in the main SOL of DIII-D associated with divertor detachment vs attachment and closure vs openness

Edge plasma code SOLPS modeling has been used to assess the effects of divertor detachment vs attachment and closure vs openness on the main scrape-off layer (SOL) plasma in DIII-D. The differences between the inner divertor and outer divertor, and the effects of divertor closure on the outer-midplane SOL plasma, have been analysed. It is found that the divertor shape and plasma state have significant influence on the SOL and pedestal plasma. (i) The inner divertor has a more significant effect on the high-field side (HFS) SOL than the outer divertor does on the low-field side (LFS) SOL, and for increasing main plasma density the ionization region in the HFS moves gradually from the divertor target into the HFS SOL, forming a high-density front in the high density discharge case. Also, particle flux reversal occurs near the separatrix. (ii) It is confirmed that divertor closure has a significant effect on the main SOL and pedestal plasma. Comparing an open and a closed divertor, it is found that for the closed divertor the ne radial profile at the OMP is higher outside the separatrix and lower inside it, due to the combined effects of the D+ and C impurity ion density. Moreover, the open divertor has higher core fueling than the closed divertor, and the closed divertor has better fuel and impurity neutral screening than the open divertor. (iii) The in/out asymmetry is also influenced by the divertor closure, increase of the closure of the outer divertor can substantially reduce the asymmetry. The closed divertor also has much higher HFS SOL density than that of the LFS SOL. These results highlight the strong dependence of neutral/impurity screening and main SOL properties on divertor geometry and plasma regimes, which needs to be taken into account in advanced divertor design for fusion reactors.

Simulations of Argon plasmas in the linear plasma device GyM with the SOLPS-ITER code

Edge plasma codes, such as SOLPS, are widely used to study plasma transport in tokamaks scrape-off layers (SOL). The possibility to apply these codes to non hydrogenic plasmas and to linear plasma devices is gaining the interests of the fusion community. These facilities play a pivotal role in plasma-material interaction studies for future fusion devices and may allow to test the code capabilities both in terms of geometry and simulated plasma species. In this contribution, we apply the SOLPS-ITER code for the simulations of Argon plasmas in the medium-flux linear machine GyM. A sensitivity scan over the pumping efficiency, transport coefficients and absorbed electron power was done, performing B2.5-EIRENE coupled simulations. A quantitative analysis of the different flux contributions is provided through the use of a two-point modelling. Comparison with experiments shows a promising qualitative and quantitative agreement, with the sole exception of the simulated neutrals pressure. Dedicated EIRENE standalone simulations were performed to investigate this issue, also highlighting the role of neutral–neutral elastic collisions at high values of the puffing intensity.

Time-dependent modeling of coupled plasma-wall dynamics

Effects of dynamic wall outgassing on divertor plasma during an edge localized mode (ELM)-like heat pulse are modeled using the newly coupled edge plasma code UEDGE and wall reaction-diffusion code FACE. Different divertor regimes are simulated depending on the degree of the plasma detachment. It is shown that hydrogen outgassing from the divertor target plate induced by ELM pulse in the semi-detached regime can trigger deep plasma detachment. The induced detachment does not spontaneously reverse after the pulse due to ineffective implantation of low energy plasma hydrogen in the target material. We also show that plasma neutrals play an essential role in heat transport in a detached divertor. The modeled hydrogen dynamics in the wall material demonstrates that the heat pulse leads to enhanced hydrogen transport toward the wall bulk.

Exploring SF- in-out asymmetry and detachment bifurcation in HL-2M with E × B by SOLPS

Abstract In this paper, the SF- (with the second X-point located near the outer target surface) in-out asymmetry and detachment bifurcation in HL-2M were explored by the edge plasma code SOLPS5.1 including electric but neglecting magnetic drifts. The modeling results reveal that the radial electric drift component (Epol × B) can play a key role on inducing the SF- in-out asymmetry due to the effective carbon (C) screen capacity and high heat dissipation. The Epol × B drift can enhance the poloidal electric drift component (Er × B), and their synergistic effects can induce a marked in-out asymmetry. Besides, the results showed that the detachment bifurcation in SF- also could occur under given conditions. The presence of four factors (including E × B, C impurity, normal-Bt, and relatively small radial anomalous transport coefficients) forms the necessary conditions for producing the detachment bifurcation. The root reason for the detachment bifurcation formation is due to the strong interaction of Te, E×B, and C impurity radiation loss in the outer divertor region with normal-Bt direction. The detachment bifurcation could be removed or mitigated by weakening their interaction. Finally, the time-dependent detachment bifurcation process is simulated in this work to better understand its physical picture.

Impact of an alternative divertor configuration on plasma detachment: pure deuterium simulations using the SOLEDGE2D-EIRENE edge transport code for HL-2M scenarios

The SOLEDGE-EIRENE edge plasma code provides solutions for particle and energy transport in the plasma edge within complex and realistic 2D geometries (Bufferand et al 2015 Nucl. Fusion 55 053025). In this work, divertor detachment is simulated on HL-2M alternative magnetic configurations in pure deuterium plasma. Starting from a typical HL-2M low single-null configuration, the snowflake plus (SF+) and snowflake minus (SF−) configurations have then been investigated. Detachment of the outer target is studied in these configurations during plasma density ramps controlled by a fueling source, with constant input power and constant radial transport coefficients. Some typical characteristics of detachment, like threshold, depth and upstream window of detachment are investigated. In the three geometries, detachment onset and evolution with upstream plasma density is characterized by the gradual displacement of a radiation front from the outer target to the main X-point, as observed in experiments. It is found that, whatever the detachment in terms of particle, momentum or power dissipation, the detachment threshold is dominated primarily by the geometrical structure of the divertor plate and does not exhibit dependence on the magnetic configuration of the diverted plasma volume. In particular, the parallel connection length in the divertor is not found to affect the detachment threshold, in contrast with simple expectations from the two-point model, but in agreement with experimental findings.

Numerical simulation of lithium granule injection in tokamak edge plasma

To understand how a Li pellet affects the background plasma, the edge plasma code SOLPS combined with a neutral-gas-shield ablation model has been employed to simulate the transport of Li species in the edge plasma during Li pellet injections in a typical EAST H-mode plasma. The simulation results show that a Li pellet 0.5 mm with velocity ∼100 m/s can penetrate the separatrix, and that the electron temperature drops quickly where the pellet passes by while the electron density needs a longer time to respond. The plasma pressure in the pedestal region grows significantly with the pellet size; the pressure gradient changes even more drastically. This work indicates that without accumulative contributions, a pellet 0.7 mm or above in diameter at an injection velocity of 100 m/s is likely to provide a pressure perturbation enough to trigger ELMs in the typical EAST H-mode plasmas used in this work.

Role of E × B on in–out divertor asymmetry in high recycling/partial detachment regimes under L-mode and H-mode conditions

The role of the E × B electric drift on background plasma and carbon impurity in–out divertor asymmetry was estimated under L-mode and H-mode conditions in the high recycling regime and partial detachment regime by using the edge plasma code SOLPS5.1. It was found that the poloidal electric drift Er × B also may play a dominant role during H-mode discharge in high recycling regime, instead of the radial electric drift Eθ × B. Moreover, it also was found that during H-mode with partial detachment both components can play simultaneously a crucial role in inducing the in–out asymmetry. Their synergistic effect can make the asymmetry much more obvious than that with either of them separately. However, Eθ × B in partial detachment during L-mode can play a main role in inducing in–out asymmetry, rather than Er × B. Besides, the role of E × B components on carbon (C) impurity in–out asymmetry was also addressed. Simulation results reveal that Er × B or Eθ × B individually have a very small effect on C impurity ions in–out asymmetry, especially Eθ × B, while their synergistic effect makes the impurity ions exhibit a much more remarkable in–out asymmetry. Moreover, it was found that the Er × B and Eθ × B drift flows in the private flux region could play a crucial role in inducing C impurity in–out asymmetry, rather than the parallel flow or electric drift flow in the upstream SOL region when only considering E × B.

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak (EAST) L-mode and ELM-free H-mode plasmas. The divertor power footprint widths, which consist of the scrape-off layer (SOL) width λq and heat spreading S, are important physical parameters for edge plasmas. In this work, a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip. Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of and Similar trends have also been demonstrated in the study of heat spreading with and In addition, studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current. The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor (CFETR).

SOLPS5.1 analysis of detachment with drifts and gas pumping effects in EAST

The aim of this paper is to estimate the effects of usual drifts and gas puffing/pumping locations on divertor detachment and Ar ion transport in the Experimental Advanced Superconducting Tokamak (EAST) by using the edge plasma code package SOLPS5.1. The simulated results reveal that which target plate first detaches depends strongly on the usual drifts, but not on the location of impurity gas puffing, which could be one of the possible explanations for the experimentally observed phenomenon (Chen et al 2013 Phys. Plasmas 20 022311) that the lower inner target first detached compared to the lower outer target with the lower outer gas puffing. The physics behind this phenomenon is that drifts not only can induce background ion flux, plasma density and temperature redistribution in the scrape-off layer (SOL) and divertor region, but also can change the Ar impurity force balance leading to Ar ions being dragged from bottom to top. Furthermore, the simulated results illustrate that the Ar ion transport in the SOL and divertor region is similar for different gas puffing locations including upstream and divertor region before partial detachment. However, the Ar ions penetrate into the core more easily, giving rise to more discharge disruption during complete detachment with upstream gas puffing than with divertor region puffing. Finally, we also estimate the effect of gas pumping on the detachment in order to realize long-pulse partial detachment in EAST. The results indicate that long-pulse partial detachment could be obtained by improving the pumping speed to match the puffing speed in case the excess Ar atoms accumulate in the core plasma during partial detachment in EAST.

A Model of Self‐similar Radiative Transfer in Resonance Lines for Testing the Edge Plasma Codes

AbstractPossibilities of obtaining the analytical solutions of the one‐dimensional equation of radiative transfer in resonance atomic/ionic lines in the Biberman‐Holstein model for testing the edge plasma codes are investigated. It is shown that for some types of similarity of spatial profiles of three characteristics, namely, background plasma density, line shape width and non‐radiation source of atomic excitation, the profile of excited atoms density may be described analytically in terms of the similarity of the above‐mentioned profiles. (© 2015 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Automated divertor target design by adjoint shape sensitivity analysis and a one-shot method

As magnetic confinement fusion progresses towards the development of first reactor-scale devices, computational tokamak divertor design is a topic of high priority. Presently, edge plasma codes are used in a forward approach, where magnetic field and divertor geometry are manually adjusted to meet design requirements. Due to the complex edge plasma flows and large number of design variables, this method is computationally very demanding. On the other hand, efficient optimization-based design strategies have been developed in computational aerodynamics and fluid mechanics. Such an optimization approach to divertor target shape design is elaborated in the present paper. A general formulation of the design problems is given, and conditions characterizing the optimal designs are formulated. Using a continuous adjoint framework, design sensitivities can be computed at a cost of only two edge plasma simulations, independent of the number of design variables. Furthermore, by using a one-shot method the entire optimization problem can be solved at an equivalent cost of only a few forward simulations. The methodology is applied to target shape design for uniform power load, in simplified edge plasma geometry.

Numerical simulation of the energy deposition evolution on divertor target during type-III ELMy H-mode in EAST using SOLPS

Impacts of particle and energy fluxes during edge localized modes (ELMs) on the divertor targets were particularly studied through Langmuir Probes in EAST (Wang et al., 2012); however, no attempt has been made to model the time-dependent ELMy H-mode of EAST yet by the edge plasma code package SOLPS. This paper aims to model the type-III ELMy H-mode discharge in EAST using SOLPS. Firstly, we adjust the perpendicular anomalous transport coefficients (PATCs) by matching the experimental upstream radial electron density and temperature profiles under given type-III ELMy H-mode discharge conditions (shot #33266) to obtain the steady-state H-mode, and then, ELMs are modeled by periodically enhancing PATCs with the parameters, such as the repetition frequency and the energy expelled from the core plasma, taken directly from the experimental data of the given EAST discharge. In this way, many experimentally inaccessible upstream parameters can be evaluated through the simulation; meanwhile, many input parameters can be provided from such simulations to other codes for understanding the damage of plasma-facing materials caused by plasma irradiation.

Simulation of hydrogen isotope retention in the gap of tungsten divertor tiles

The use of castellated divertor structures in fusion devices may increase the amount of hydrogen isotope retention because of the presence of gaps between the tiles. To address this issue, a two dimensional Hydrogen Isotope Inventory Processes code (HIIPC2D) is developed. To represent the effect of the gap geometry on the incident plasma flux reasonably, a two dimensional kinetic edge plasma code is integrated into HIIPC2D. The simulation results indicate that the existence of gaps of divertor tiles does increase the HI retention in the short term, but in the long term its contribution to the total HI retention will be small.

Modelling of edge plasma transport during H-mode of EAST by SOLPS5.0

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.

Modeling of Local Edge Plasma Perturbations Induced by a Biased Probe

AbstractThe Langmuir probe is a standard diagnostic tool for laboratory plasmas, routinely used for edge plasmas in magnetic fusion devices. However, recent studies on the Alcator C‐Mod tokamak indicate the possibility of a substantial local plasma perturbation by a negatively biased divertor plate target probe under high recycling conditions when the electron temperature is under ∼10 eV. The effects of the probe perturbation of the plasma are studied using the tokamak edge plasma code UEDGE for parameters relevant to Alcator C‐Mod edge. It is found that at sufficiently negative bias voltage, the probe substantially modifies the local plasma characteristics. In particular, this is relevant to the experimentally known “death‐ray” phenomenon, where the plasma pressure near the probe strongly exceeds the upstream value. Certain features of the “death‐ray” phenomenon are reproduced in the modeling, pointing to interplay of plasma and neutrals transport, atomic physics, and sheath heat‐flux boundary conditions (© 2012 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

New approach in two-dimensional fluid modeling of edge plasma transport with high intermittency due to blobs and edge localized modes

A new approach is proposed to simulate intermittent, non-diffusive plasma transport (via blobs and filaments of edge localized modes (ELMs)) observed in the tokamak edge region within the framework of two-dimensional transport codes. This approach combines the inherently three-dimensional filamentary structures associated with an ensemble of blobs into a macro-blob in the two-dimensional poloidal cross-section and advects the macro-blob ballistically across the magnetic field, B. Intermittent transport is represented as a sequence of macro-blobs appropriately seeded in the edge plasma according to experimental statistics. In this case, the code is capable of reproducing both the long-scale temporal evolution of the background plasma and the fast spatiotemporal dynamics of blobs. We report the results from a two-dimensional edge plasma code modeling of a single macro-blob dynamics, and its interaction with initially stationary background plasma as well as with material surfaces. The mechanisms of edge plasma particle and energy losses from macro-blobs are analyzed. The effects of macro-blob sizes and advection velocity on edge plasma profiles are studied. The macro-blob impact on power loading and sputtering rates on the chamber wall and on inner and outer divertor plates is discussed. Temporal evolution of particle inventory of the edge plasma perturbed by macro-blobs is analyzed. Application of macro-blobs to ELM modeling is highlighted.