Divertor Transport Research Articles

He II line intensity measurements in the JET tokamak

Abstract An understanding of the behaviour of the D or He fuel used in tokamak discharges is essential in analyses such as modelling edge and divertor transport and the erosion of the vessel walls. However, poor agreement is found between measurements made on the JET tokamak and collisional-radiative models used to predict the hydrogen-like D and He line intensities. The range of temperatures of the plasmas emitting the radiation is also limited, in contrast to that for many impurities for which a wide range is possible. This is particularly so for He II whose line intensities tend to have the same near-constant ratios in most pulses, suggesting that the emission originates in plasma regions with very similar electron temperatures. To gain understanding and to allow quantitative comparisons with theoretical models, extensive observations of the VUV Lyman series have been made, for all discharge scenarios run during three He campaigns. Those for He discharges in both JET ITER-like wall (JET-ILW) and JET C (JET-C) campaigns are presented here. He discharges have the advantage of fewer impurities resulting in less complex spectra than when D is used as the fuel. However, the characteristics of the observed discrepancies are similar in both species, allowing He to be used as a proxy for D in order to gain understanding of the discrepancy. In addition, the study of He avoids the complication of molecular species contributing to the level populations. Opacity effects are also expected to be less severe in He discharges. Nevertheless, so as to ensure that the measurements are not unduly affected by opacity, comparisons have also been made with emission from Balmer and Paschen series members. Measurements of both line intensities and their ratios are presented for all-pulse surveys and for individual pulses. In exceptional cases in which the He emission is intense a dependence on the He II line intensity is demonstrated. The discrepancy between these measurements and the theoretical models is illustrated.

Characteristics of the SOL ion-to-electron temperature ratio on the J-TEXT tokamak with different plasma configurations

Accurate measurement of the average plasma parameters in the edge region, including the temperature and density of electrons and ions, is critical for understanding the characteristics of the scrape-off layer (SOL) and divertor plasma transport in magnetically confined fusion research. On the J-TEXT tokamak, a multi-channel retarding field analyzer (RFA) probe has been developed to study average plasma parameters in the edge region under various poloidal divertor and island divertor configurations. The edge radial profile of the ion-to-electron temperature ratio, τ i/e, has been determined, which gradually decreases as the SOL ion self-collisionality, , increases. This is broadly consistent with what has been observed previously from various tokamak experiments. However, the comparison of experimental results under different configurations shows that in the poloidal divertor configuration, even under the same , τ i/e in the SOL region becomes smaller as the distance from the X-point to the target plate increases. In the island divertor configuration, τ i/e near the O-point is higher than that near the X-point at the same , and both are higher than those in the limiter configuration. These results suggest that the magnetic configuration plays a critical role in the energy distributions between electrons and ions at the plasma boundary.

Modeling of plasma beta effects on the island divertor transport in the standard configuration of W7-X

The influence of plasma beta effects on the edge plasma transport in the Wendelstein 7-X standard configuration is studied systematically by using EMC3-EIRENE combined with a 3D equilibrium code named HINT. The magnetic topology changes induced by plasma beta effects are significantly reflected in plasma transport behaviors and heat flux patterns on divertor targets. After validating the modeling strategy by comparisons with experimental data, the extended simulations for high performance plasmas show that the threshold separatrix density for accessing the power detachment is reduced in higher beta plasmas. Compared with the vacuum field case, the impurity radiation distributions with finite beta effects are modified in the magnetic island region. The divertor heat flux is distributed more evenly along the toroidal direction on the strike line at the vertical target. The strike line on the horizontal target moves towards the pumping gap with an increase in the plasma beta. In addition, the different pressure profiles with the same central beta also result in a modified heat flux pattern on the divertor targets.

Multi-code estimation of DTT edge transport parameters

The main goal of the Divertor Tokamak Test facility (DTT) is to operate with a high value of power-exhaust-relevant parameter P S O L /R in plasma scenarios similar to those foreseen for the Demonstration Fusion Power Plant (DEMO) in terms of low collisionality and neutral opacity. For these unique characteristics, accurate modelling of the principal scenario is necessary for machine designing. In edge numerical codes, cross-field transport profiles have a high impact on modelling results. This work aims at providing a coherent set of transport parameters for DTT full-power (FP) single-null (SN) scenario edge modelling. To evaluate such parameters for DTT, a transport analysis on the current machine has been performed using SOLEDGE2D-EIRENE and SOLPS-ITER. The transport parameters to be used in the simulations of the DTT single-null scenario were selected using two complementary methods. The first is the modelling of JET and Alcator C-Mod (C-Mod) with SOLEDGE2D-EIRENE and SOLPS-ITER, validating transport parameters by comparing modelling results to experimental data from pulses which are considered DTT-relevant . JET pulses were selected with the highest auxiliary input power (from 29 to 33 MW), plasma current and toroidal field to better match DTT parameters; nitrogen and neon seeded pulses were selected to check possible seeding material dependencies. The considered C-Mod pulse better matches DTT plasma density and neutral opacity. Transport parameters are then scaled to DTT according to scaling laws. The second method derives the transport parameters by tuning their values inside the DTT separatrix to reproduce the pedestal profiles predicted by the EPED model via the Europed code and applied in DTT. The derived set of DTT transport parameters is consistent with the results obtained by modelling present machines, reproduces the expected heat flux decay length in detached conditions and, inside the separatrix, reproduces the predicted pedestal using transport parameters which are coherent with what is predicted by the quasi-linear turbulent model QuaLiKiz. • Edge and divertor transport parameters of the DTT single-null scenario were studied. • The edge transport parameters of JET and C-Mod pulses were derived by SOLEDGE2D-EIRENE and SOLPS modelling. • Transport parameters were scaled from present machines to DTT. • First principles core modelling with JETTO confirms the validity of the previous estimations.

Power exhaust and core-divertor compatibility of the baffled snowflake divertor in TCV

A baffled snowflake minus low-field side (SF-LFS) is geometrically-optimised in tokamak à configuration variable, increasing divertor neutral pressure, to evaluate the roles of divertor closure (comparing with an unbaffled SF-LFS) and magnetic geometry (comparing with a baffled single null (SN)) in power exhaust and core-divertor compatibility. Ohmically-heated L-mode discharges in deuterium, with a line-averaged core density of approximately m−3, are seeded with nitrogen to approach detached conditions. Baffles in the SF-LFS configuration are found to reduce the peak outer target heat flux by up to , without significantly affecting the location of the inter-null radiation region or the core-divertor compatibility. When compared to the baffled SN, the baffled SF-LFS exhibits a reduction in the outer target heat flux by up to and the ability to balance the strike-point distribution of heat flux. These benefits are less significant with N2 seeding, with similar peak target quantities (such as heat flux, electron temperature and ion flux) and divertor radiated power. Despite a radiating region located farther from the confined plasma for the SF-LFS than the baffled SN, no change in core confinement is observed. Core effective charge even indicates an increase in core impurity penetration for the SF-LFS. These experiments constitute a good reference for detailed model validations and extrapolations, exploring important physics such as core impurity shielding and the dependence of divertor cross-field transport on magnetic geometry.

Control of tungsten impurity source and edge transport using different gas injection with full tungsten divertor on EAST

• With the newly installed EAST lower W divertor, W source and edge transport are experimentally investigated for different Ne and D 2 gas injection conditions. • W erosion on the whole divertor target can be well suppressed or mitigated by suitable edge Ne or D 2 injection when divertor detachment is achieved. • A W screening enhancement is obtained during the attached divertor condition due to D 2 injection. • After the divertor detachment, the core W content is observed to be kept in a high level, although W source from the divertor is strongly reduced. Tungsten (W) erosion and edge transport are investigated for EAST L-mode discharges with different gas injection. It is found that W erosion can be suppressed or mitigated by Ne or D 2 seeding when divertor detachment is achieved. Compared to edge D 2 fueling, Ne seeding from the divertor target is favorable for full detachment condition and thus W erosion suppression. Increasing the upstream plasma density by edge D 2 fueling can affect the divertor condition, which may lead to a W erosion mitigation. D 2 puffing at divertor target is less effective on increasing the upstream plasma density than OMP D 2 puffing, and thus it is less effective on reduction of W erosion rates. W gross erosion profiles with different amount of injected D 2 at the divertor are reproduced by a mixed material W erosion model, which indicates that there exist a Li-C overlayer on the W surface of EAST divertor. Ne and D 2 injection are also found to have different impact on the normalized core W density. For the attached divertor condition, divertor Ne seeding will increase W leakage, but a suitable D 2 fueling from divertor target can strengthen the edge W screening. After the divertor detachment, although the W source is dramatically reduced, the W core density is kept in a high level for the Ne seeding discharges, and even increased for the D 2 fueling discharges.

Modeling of deuterium and carbon radiation transport in MAST-U tokamak advanced divertors

Modest effects of deuterium and carbon radiation opacity in the Super-X and snowflake divertor plasmas are predicted for MAST Upgrade tokamak with core plasma input power 2.5–5 MW and plasma current 1 MA. The radiation transport modeling is based on the SOLPS-EIRENE and UEDGE code divertor plasma predictions. Two radiation transport models are used: one is based on a full radiation transport equation implemented in the radiation transport and collisional-radiative code CRETIN (without feedback on the background plasma), and another is an internal self-consistent UEDGE model with ionization, recombination, and heating rates corrected for Ly α line trapping based on the escape probability model implemented in CRETIN. In MAST-U, the Super-X and snowflake divertor plasmas are predicted to reach detached regimes at lower upstream densities than the standard divertor, and the conclusion still holds with radiation transport effects included. At neutral densities m−3, modest Ly α deuterium line trapping with optical depths 10–15 is predicted in the Super-X divertor. Divertor plasmas are optically thin to other Lyman and Balmer lines, as well as to strong C III and C IV lines that are responsible for most of divertor radiated power. Insignificant changes (within a few percent) to divertor deuterium ionization and recombination rates are found. Radiation fluxes on outer divertor target are modified within a factor of 2–3 when the radiation transport is accounted for, and a similar variation is found due to the line shape models that define the absorption and emission line profiles in the radiation transport modeling. The predicted Lyman and Balmer spectral intensities are significantly modified due to radiation trapping. A measurement of divertor radiation transport effects is discussed using the Ly β /Ba α line ratio. In the snowflake divertor configuration, divertor plasmas are found to be optically thin to Lyman series lines within a large range of parameter variations that include upstream density, divertor transport coefficients, and magnetic configurations. Modest radiation transport effects are only found in a few cases with strongest divertor transport and magnetic configurations closest to the ideal snowflake configuration, however, the plasma background models that were used are yet to be validated with an experiment.

Progress in DIII-D towards validating divertor power exhaust predictions

UEDGE simulations highlight the role of cross-field drifts on the onset of detached conditions, and new calibrated divertor vacuum ultra violet (VUV) spectroscopy is used to challenge the predictions of radiative constituents in these simulations. UEDGE simulations for DIII-D H-mode plasmas with the open divertor with the ion ∇B-drift towards the X-point show a bifurcated onset of the low field side (LFS) divertor detachment, consistent with experimentally observed step-like detachment onset (Jaervinen A.E. et al 2018 Phys. Rev. Lett. 121 075001). The divertor plasma in the simulations exhibits hysteresis in upstream separatrix density between attached and detached solution branches. Reducing the drift magnitude by a factor of 3 eliminates the step-like detachment onset in the simulations, confirming the strong role of drifts in the bifurcated detachment onset. When measured local plasma densities and temperatures are within proximity of predicted values in the simulations, there is no shortfall of the local emission of the dominant resonant radiating lines. However, the simulations systematically predict a factor of two lower total integrated radiated power than measured by the bolometer with the difference lost through radial heat flow out of the computational domain. Even though there is no shortfall in the emission of the dominant lines, a shortfall of total radiated power can be caused by underpredicted spatial extent of the radiation front, indicating a potential upstream or divertor transport physics origin for the radiation shortfall, or shortfall of radiated power in the spectrum between the dominant lines. In addition to the underpredicted spatial extent, in detached conditions, the simulations overpredict the peak radiation and dominant carbon lines near the X-point, which can be alleviated by manually increasing divertor diffusivity in the simulations, highlighting the ad hoc cross-field transport as one of the key limitations of the predictive capability of these divertor fluid codes.

Developing physics basis for the snowflake divertor in the DIII-D tokamak

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power –5 MW and a range of plasma currents MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10–15% of total outer divertor power. Measurements of electron pressure and poloidal beta support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4–4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks.

Impact of 3D magnetic field structure on boundary and divertor plasmas in stellarator/heliotron devices

This paper overviews recent progress on the experimental identification and physics interpretation of 3D effects of magnetic field geometry on divertor transport. The 3D effects are elucidated as a consequence of competition between transports parallel (||) and perpendicular (⊥) to magnetic field, in open field lines cut by divertor plates, or in magnetic islands. The competition has strong impacts on divertor functions, such as determination of density regime, impurity screening, and detachment control. The effects of magnetic perturbation on the edge electric field and turbulent transport are also discussed. Based on the experiments and numerical simulations, key parameters governing the 3D transport physics for the individual divertor functions, e.g. pumping efficiency through divertor density regime, impurity screening and detachment control, are discussed.

Analysis of rotating collectors from the private region of JET with carbon wall and metallic ITER-like wall

Rotating collectors are used in JET to provide time-resolved measurements of erosion and redeposition of vessel materials. The silicon collecting discs rotate behind an aperture, driven by pulsing of the toroidal magnetic field, with the deposits analysed ex-situ by nuclear reaction analysis. The angular dependence of deposition is mapped to discharge number using the B-field history, allowing the influence of different plasma configurations and parameters to be investigated. A simple geometrical model using sputtering and reflection from the strike point has qualitatively reproduced the deposition found on collectors located under the central divertor tile and facing towards the inner strike point. The beryllium deposition on the ITER-like wall (ILW) collector showed an order of magnitude reduction in deposition compared to carbon deposition on the JET-C collector. This decreased deposition is attributed to low long range divertor transport due to reduced chemical sputtering/erosion and codeposition of beryllium relative to carbon.

Simulation of tokamak SOL and divertor region including heat flux mitigation by gas puffing

Two-dimensional (2D), scrape-off layer (SOL)-divertor transport simulations are performed using the integrated plasma-neutral-impurity code KTRAN developed at Seoul National University. Firstly, the code is applied to reproduce a National Spherical Torus eXperiment (NSTX) discharge by using the prescribed transport coefficients and the boundary conditions obtained from the experiment. The plasma density, the heat flux on the divertor plate, and the D α emission rate profiles from the numerical simulation are found to follow experimental trends qualitatively. Secondly, predictive simulations are carried out for the baseline operation mode in Korea Superconducting Tokamak Advanced Research (KSTAR) to predict the heat flux on the divertor target plates. The stationary peak heat flux in the KSTAR baseline operation mode is expected to be 6.5 MW/m2 in the case of an orthogonal divertor. To study the mitigation of the heat flux, we investigated the puffing effects of deuterium and argon gases. The puffing position is assumed to be in front of the strike point at the outer lower divertor plate. In the simulations, mitigation of the peak heat flux at the divertor target plates is found to occur when the gas puffing rate exceeds certain values, ∼1.0 × 1020 /s and ∼5.0 × 1018 /s for deuterium and argon, respectively. Multi-charged impurity transport is also investigated for both NSTX and KSTAR SOL and divertor regions.

Divertor electron temperature and impurity diffusion measurements with a spectrally resolved imaging radiometer

A divertor imaging radiometer (DIR) diagnostic is being studied to measure spatially and spectrally resolved radiated power P(rad)(λ) in the tokamak divertor. A dual transmission grating design, with extreme ultraviolet (~20-200 Å) and vacuum ultraviolet (~200-2000 Å) gratings placed side-by-side, can produce coarse spectral resolution over a broad wavelength range covering emission from impurities over a wide temperature range. The DIR can thus be used to evaluate the separate P(rad) contributions from different ion species and charge states. Additionally, synthetic spectra from divertor simulations can be fit to P(rad)(λ) measurements, providing a powerful code validation tool that can also be used to estimate electron divertor temperature and impurity transport.

Simulation Study of L/H Transition with Self‐Consistent Integrated Modelling of Core and SOL/Divertor Transport

AbstractWe have developed a self‐consistent integrated simulation model of core and scrape‐off‐layer (SOL)/divertor transport. This model enables the user to investigate operational scenarios with compatible high‐confinement core plasmas and detached divertor plasmas. A 1.5D core code TOPICS‐IB and a 2D divertor code SONIC are successfully coupled using a Multiple Program Multiple Data parallel computing system. This integrated code, which can include Monte‐Carlo models for neutrals and impurities, provides efficient plasma/neutral simulations. The dynamic simulation for the L/H transition in JT‐60SA is carried out by this integrated code using a current‐diffusive ballooning mode (CDBM) transport model that includes the E×B shearing effect. The impact of SOL/divertor transport on the L/H transition is studied (© 2012 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Coherence Imaging of Flows in the DIII‐D Divertor

AbstractVarious spatial heterodyne polarization interferometers for spectrally‐resolved optical imaging of edge and core parameters in high temperature magnetized plasmas are described. Applications for such “coherence imaging” (CI) systems include imaging motional Stark effect and Zeeman effect polarimetry for determination of the magnetic field pitch angle, and passive and active (charge exchange recombination spectroscopy ‐ CXRS) Doppler imaging of plasma temperature and flow. In this paper we describe spatial heterodyne coherence imaging systems and present first results of Doppler flow imaging in the DIII‐D divertor.Instruments have been installed for imaging flows in the divertor and scrape‐off‐layer in the DIII‐D tokamak and also for Doppler imaging on the H‐1 heliac [1]. In the former case, single snapshot interferometric images of the plasma in CII 514nm, and CIII 465nm emission have been demodulated to obtain flow and ion temperature projections in both the scrape‐off‐layer and divertor. Flow field amplitudes in the divertor are found to be broad agreement with UEDGE modeling [2], and point the way towards experiments that address important divertor transport issues in future (© 2011 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

A study of tungsten migration in the Alcator C-Mod divertor

A toroidally-continuous row of bulk tungsten (W) tiles was installed near the typical location of the outer strike-point in the Alcator C-Mod divertor and was removed for analysis after two campaigns (3151 plasma discharges). External beam (in-air) particle induced X-ray emission (PIXE) analysis has been used to observe campaign-integrated W migration patterns in the otherwise Mo divertor using the W as a tracer for divertor erosion and transport. With a set of high spatial resolution PIXE measurements (∼3mm resolution), poloidal profiles of W migration have been constructed. Redeposition of up to 1.5×1021atoms/m2 (∼20nm equivalent thickness) is observed on the outer and inner divertor, and in the private-flux region. The campaign-integrated deposition on the analyzed tiles corresponds to net W erosion of ∼60nm (>0.014nm/s). The migration profiles indicate sputtering and prompt redeposition in the outer divertor, neutral and ion transport through the private–flux region, and transport in the scrape off layer.

Comparison between stellarator and tokamak divertor transport

This paper compares the essential divertor transport features of the poloidal divertor, which is well developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences on the functioning of a divertor. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modelling using the EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations.

Comparative divertor-transport study for helical devices

Using the island divertors (IDs) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine sizes following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. The fundamental role of low-order magnetic islands in both divertor concepts is emphasized. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime, which is absent from W7-AS and LHD, is predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-ß plasmas.

Analysis on EAST LHCD Operation Space by Using Simple Core-SOL-Divertor Model

A simple core-SOL-divertor model (CSD model) was developed to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. In the CSD model, the core plasma model of ITER physics guidelines and the two-point SOL-divertor model are applied. This CSD model is validated by the two dimensional divertor transport code (B2-EIRINE) and by the JT-60U divertor recycling database, and this model is applicable to the low- and high-recycling state of the divertor plasma. The CSD model is applied to the study of the EAST operational space with lower hybrid current drive under various kinds of trade-off for the basic plasma parameters, and the relationship between the operational space and the plasma discharge duration is also discussed.

Erosion and deposition in the JET MkII-SRP divertor

Carbon-13 labelled methane was injected into the outer divertor during a series of H-mode discharges on the last day of operations with the JET MkII-SRP divertor. Tiles from around the vessel were removed during the subsequent shutdown and surface deposits were analysed by IBA techniques and SIMS. First attempts to model the pattern of 13C deposition using EDGE2D are reported. Erosion of W markers at the outer divertor was observed, with implications for the ITER-like wall experiment planned for JET, whilst thin film growth in the same region has been followed by the effect on infra-red measurements. The composition of thick films deposited at the inner divertor during the MkII-SRP campaign, and the migration to the inner corner of the divertor observed by a quartz micro-balance, provide further information on divertor transport.