ELMy H-mode Plasmas Research Articles

Construction and experimental tests of the high frequency pellet injection system on EAST tokamak

A high frequency deuterium pellet injection system based on extrusion and pneumatic acceleration has been recently developed and built for EAST tokamak, which is one of fully superconducting Tokamaks with ITER-like plasma configuration. It is mainly used for plasma edge localized mode control and can achieve 1–50 Hz injection rate. It has been completely installed on EAST up to now. A new differential pumping system and pellet transmission lines have been designed according to the actual condition of EAST. Recently we have tested the system systematically and injected multiple pellets into type-I ELMy H-mode plasma on EAST for the first time. The test and experimental results show that the system can inject continuous pellets into plasma stably, and ELMs can be triggered by deuterium pellets in H-mode plasma with negligible impact on density and stored energy. And it was also found that ELM trigger efficiency is less than 100 % sometimes. This lays an important foundation for further experimental study of ELM modulation using deuterium pellet injection on EAST.

Scaling study for positive magnetic shear ELMy H-mode plasmas in JT-60U

Scaling of the density peaking for positive magnetic shear ELMy H-mode plasmas in JT-60U has been developed. Although the density peaking generally depends on collisionality, a variation of the density peaking factor for the same collisionality exists, and the variation is different between plasmas with co-directed and ctr-directed toroidal rotation(co-VT plasmas and ctr-VT plasmas). The variation of the co-VT plasmas can be explained by particle source profile. As a result of scaling, the peaking factor of the co-VT plasmas depends on the particle source rate profile from neutral beams (NBs) as much as collisionality. The dataset of the ctr-VT plasmas has a larger variation of the peaking factor compared to that of the co-VT plasmas. The larger variation stems from that the peaking factor of the ctr-VT plasmas also depends on the normalized ion temperature gradient at the edge region. As a result of scaling, the normalized ion temperature gradient influences the peaking factor as much as or a little larger than collisionality. The parameter regime of the ctr-VT plasmas ranges from the trapped electron mode (TEM) to the ion temperature gradient mode (ITG mode). The plasmas with the positive correlation between the normalized density gradient and the normalized ion temperature gradient are dominated by the TEM. On the other hand, the plasmas with the negative correlation are dominated by the ITG mode.

Experimental energy confinement time scaling with dimensionless parameters in C-Mod I-mode plasmas

The dependence of energy confinement time on gyroradius, , and normalized collisionality in I-mode plasmas is investigated through dedicated C-Mod experiments scanning dimensionless parameters with constant safety factor. The gyroradius scaling is calculated to be , which suggests core transport may scale with gyro–Bohm physics, indicating favorable extrapolation of the I-mode regime to future devices at low . The scaling exponent for is calculated to be small, but positive (), and the exponent for is deemed inconclusive () due to high correlation with the other two dimensionless variables in the dataset, and therefore requires further investigation. The individual dimensionless parameter scaling is compared to calculations from larger C-Mod I-mode datasets as well as multi-machine scaling laws for ELMy H-mode and L-mode plasmas. Multiple regression techniques and principal component analysis are used to clarify the single parameter scalings and analyze parameter significance within the dataset.

Observation of filament-like structures in ELMy H-mode plasma with a VUV imaging system developed on the EAST tokamak

On the EAST tokamak, filament-like structures have been observed in ELMy H-mode discharges with a high-speed vacuum ultraviolet (VUV) imaging system. The topos, chronos and their weight can be obtained simultaneously by performing the so-called singular value decomposition (SVD) analysis of raw VUV imaging data. The fluctuation amplitude is observed to be suppressed and enhanced gradually in the edge localized mode (ELM) crash and pedestal recovery phase in the chronos, respectively, while filament-like structures can only be found in the pedestal recovery phase on the topos. The mode structure, i.e. m/n = 36/9 (m and n are the poloidal and toroidal mode number, respectively) with = 0.95, = 0.07 ( and denote the mode location and mode width, respectively) is derived by a comparison of the synthetic images and the experimental imaging data.

Impact of impurity mixture gas seeded by supersonic molecular beam injection on edge-localized modes in the HL-2A tokamak

Impurity seeding experiments in ELMy H-mode plasmas by using supersonic molecular beam injection (SMBI) have been performed in the HL-2A tokamak. Besides pure impurity gas (neon, argon), impurities mixed with the main ion fuelling gas by different ratios were seeded into the plasma. It has been observed that the ELM behavior and plasma confinement are very sensitive to the impurity ratio. For 30% Ne impurity seeding, large ELMs are replaced by high frequency bursts (HFBs), which have higher frequency and smaller amplitude compared to the ELMs and the peak heat load on the divertor plate was considerably reduced. Analyses indicate that HFBs originate from the pedestal region and enhance the pedestal particle transport, resulting in a softened pedestal gradient after SMBI. However, for the pure Ne and Ar SMBI seeding, the ELM frequency is decreased and the plasma confinement is moderately improved. Experimental observations suggest that the impurity mixture gas plays a role in changing the pedestal dynamics and there should be an optimal impurity ratio for efficient heat load control in ELMy H-mode plasmas.

Recent advances in EAST physics experiments in support of steady-state operation for ITER and CFETR

Since the last IAEA Fusion Energy Conference in 2016, the EAST physics experiments have been developed further in support of high-performance steady-state operation for ITER and CFETR. First demonstration of a >100 s time scale long-pulse steady-state scenario with a good plasma performance (H98(y2) ~ 1.1) and a good control of impurity and heat exhaust with the upper tungsten divertor has been achieved on EAST using the pure radio frequency (RF) power heating and current drive. The EAST operational domain has been significantly extended towards a more ITER and CFETR related high beta steady-state regime (βP ~ 2.5 and βN ~ 1.9 of using RF and NB and βP ~ 1.9 and βN ~ 1.5 of using pure RF). A large bootstrap current fraction up to 47% has been achieved with with q95 ~ 6.0–7.0. The interaction effect between the electron cyclotron resonant heating and two lower hybrid wave systems has been investigated systematically, and applied for the improvement of current drive efficiency and plasma confinement quality in the steady-state scenario development on EAST. Full edge-localized mode (ELM) suppression using the n = 2 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with a range of the edge safety factor of q95 ≈ 3.2–3.7 on EAST. Reduction of the peak heat flux on the divertor was demonstrated using the active radiation feedback control. An increase in the total heating power and improvement of the plasma confinement are expected using a 0D model prediction for a higher bootstrap fraction. Towards a long-pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active water-cooling will be installed, together with further increase and improvement of heating and current drive capability.

Deuterium pellet fueling in type-III ELMy H-mode plasmas on EAST superconducting tokamak

Type-III ELMy H-mode plasmas are a valuable operating scenario, as they may simultaneously achieve fusion power gain of Q = 10 and acceptable steady-state and peak power fluxes to the plasma-facing components. It is important to investigate cryogenic deuterium pellet injection fueling in this kind of H-mode plasma since it has been chosen as a main plasma fueling technique for ITER. Recently, pellet fueling experiments have been successfully carried out on EAST, an ITER-like tokamak. A maximum electron density of 0.73 × nGW has been obtained. It is found that ELM behavior was modified immediately and a string of small ELMs were induced when pellets were injected at a low velocity from ˜40 cm above the mid-plane on the low field side, which is different from the fact that pellets trigger signal ELMs in Type-I ELMy H-mode plasmas in other tokamaks. The decreased edge temperature caused by the cryogenic pellet was proposed as the reason for this ELM modification. These results will provide a useful reference for the pellet fueling in a Type-III ELMy H-mode plasma on ITER in the future.

Progress toward divertor detachment on TCV within H-mode operating parameters

Recent experiments on Tokamak à Configuration Variable have made significant progress toward partial detachment of the outer divertor in neutral beam heated H-mode plasmas in conventional and alternative divertor configurations. The heating power required to enter H-mode was measured in a range of divertor configurations. It is found that at the core densities most favourable for H-mode access, the L–H threshold power is largely independent of the poloidal flux expansion and major radius of the outer divertor, and in the snowflake minus configuration. A factor 2 reduction in the outer divertor power load was achieved in ELM-free (using a fuelling and nitrogen seeding) and ELMy (using nitrogen seeding) H-mode plasmas. No significant reduction in the outer divertor particle flux was achieved in the ELM-free scenarios, compared with ~30% reduction in the most strongly detached ELMy cases. The poloidal flux expansion at the outer divertor was not found to significantly alter the cooling of the divertor in the ELM-free scenarios.

Modelling of the effect of ELMs on fuel retention at the bulk W divertor of JET

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The ene ...

Millimeter-wave beam scattering by edge-plasma density fluctuations in TCV

In magnetically confined fusion devices, the use of millimeter waves (mmw) at the electron cyclotron (EC) frequencies ranges from plasma diagnostics to plasma heating, current drive and core confinement preservation. For large tokamaks such as ITER, numerical simulations and analytical estimates suggest that plasma edge turbulence could significantly broaden the EC-beam, possibly preventing tearing modes stabilization at the designed power levels. We report measurements of mmw-beam scattering by plasma turbulence in the TCV tokamak. A mmw-Gaussian beam is injected from the top of the device and the transmitted power is measured at the bottom. We show that the measured plasma density fluctuations in the upper part of the scrape-off layer (SOL) are the cause of fluctuations of the transmitted mmw-power. A full-wave model based on COMSOL multiphysics is presented and compared against the wave-kinetic-equation solver WKBeam in a TCV case. Using the SOL turbulence simulations from the GBS code, comparison between the scattering effect on the mmw-beam with both the full-wave simulations and the experiments are ongoing. We also present experimental observations of rapid changes in the transmitted power caused by ELMs in ELMy H-mode plasma.

Pedestal structure and energy confinement studies on TCV

High external gas injection rates are foreseen for future devices to reduce divertor heat loads and this can influence pedestal stability. Fusion yield has been estimated to vary as strongly as so an understanding of the underlying pedestal physics in the presence of additional fuelling and seeding is required. To address this, a database scanning plasma triangularity, fuelling and nitrogen seeding rates in neutral beam (NBH) heated ELM-y H-mode plasmas was constructed on TCV. Low nitrogen seeding was observed to increase pedestal top pressure but all other gas injection rates led to a decrease. Lower triangularity discharges were found to be less sensitive to variations in gas injection rates. No clear trend was measured between plasma top Pe and stored energy which is attributed to the non-stiffness of core plasma pressure profiles. Peeling ballooning stability analysis put these discharges close to the ideal MHD stability boundary. A constant for D in the relation pedestal width , was not found. Experimentally inferred values of D were used in EPED1 simulations and gave good agreement for pedestal width. Pedestal height agreed well for high triangularity but was overestimated for low triangularity. IPED simulations showed that relative shifts in pedestal position were contributing significantly to the pedestal height and were able to reproduce the measured profiles more accurately.

Simulations on W impurity transport in the edge of EAST H-mode plasmas

This paper provides an investigation of the production/transport properties of W impurity in the edge of experimental advanced superconducting tokamak (EAST) H-mode discharges with upper-single-null configurations by using DIVIMP Monte-Carlo code. The background plasmas are provided by SOLPS5.0 calculations. Firstly, to address the detailed dependence of W impurity behaviors on plasma conditions in the scraped-off layer (SOL)/divertor region of EAST, two pre-edge localized mode (ELM) cases with divertor plasmas in high-recycling and partially-detached regimes, respectively, are considered and compared. It is found that, due to the competitions between the thermal and friction forces (FFs), in the high-recycling case, large quantities of W impurities are transported to the upstream; while in the partially-detached case, most of the W impurities are located near the inner and outer divertors. Furthermore, the W core contamination in type-I ELMy H-mode plasmas of EAST has been estimated by the DIVIMP-SOLPS5.0 simulations, without the consideration of the release of W impurity from the core due to ELMs. During the ELM, simulations exhibit the low-recycling regime of divertor plasma, and high (several hundred eV) target plasma temperatures together with low SOL collisionality ( ∼ 1–3), lead to an order of magnitude increase in the W core contamination rate. However, in the ELM-recovering phase, with the decrease (increase) of divertor plasma temperature (density), significant increase of divertor retention is obtained due to the remarkable increase of FFs on W ions. Besides, intra-ELM phase approximately contributes to more than 50% of the total W core contamination per an ELM cycle. This work represents a step toward a deeper understanding of the W impurity production/transport in EAST H-mode discharges.

Experimental observation of reverse-sheared Alfvén eigenmodes (RSAEs) in ELMy H-mode plasma on the EAST tokamak

Reverse-sheared Alfvén eigenmodes (RSAEs) have been observed by using an interferometer and ECE diagnostics in NBI heated ELMy H-mode plasma on EAST tokamak. A typical feature of these modes is a fast frequency sweeping upward from ∼80 kHz to ∼110 kHz in hundred milliseconds during which the plasma temperature, density and rotation keeps no change. Only core channels of the interferometer can observe these modes, implying a core localized mode. The ECE measurement further showed that these modes located at about ρ = 0.37–0.46, just around the position of qmin with ρ ∼ 0.4. These core localized modes are very weak in the magnetic fluctuations measured by mirnov probes mounted at the machine vacuum vessel. A multiple frequency fluctuation component, seemingly the so-called ‘grand cascades’, was also clearly observed on the ECE signal at ρ = 0.46. During the phase, a transient internal transport barrier (ITB) in ion temperature and toroidal rotation was observed and the ITB foot was just close to the position of qmin. A modulation of RSAE frequency by ELM event was observed and this modulation could be attributed to rotation decrease or qmin increase due to ELM. Further study of these modes in EAST can provide valuable constraints for the q profile measurement and will be important for the long pulse operation.

Interaction of metal dust adhered on castellated substrates with the ELMy H-mode plasmas of ASDEX-Upgrade

Castellated substrates with adhered micron dust have been exposed in the outer ASDEX-Upgrade divertor to ELMy H-mode discharges. Beryllium proxy (chromium, copper) and refractory metal (tungsten, molybdenum) dust has been deposited on the plasma-facing and plasma-shadowed sides of the monoblocks as well as the bottom of the gaps. Interaction with time-averaged transient heat loads up to 5 MW m−2 led to dust remobilization, clustering, melting and wetting-induced coagulation. The amount of dust released in the vessel has been quantified and remobilized dust trajectories inferred. Gaps can efficiently trap locally adhered dust, but dust detaching from adjacent monoblocks does not preferentially move inside the gaps implying that they do not constitute a dust accumulation site. Heat transfer simulations of melting events are also reported taking into account heat constriction due to the finite contact area and the presence of surface roughness.

On the control system preparation for ELM pacing with vertical kicks experiments at TCV

High confinement plasma regimes (H-mode) are the scenarios with the highest fusion rates and the operating regime foreseen for the ITER scenario with QDT = 10. The steep gradients of density and temperature drive magnetohydrodynamic (MHD) edge localized modes (ELMs). Although ELMs are beneficial for controlling particle exhaust and plasma impurities, the excessive energy they deposit on the plasma facing components may lead to erosion or melting, reducing the components life cycle. By controlling the ELM frequency in an ELMy H-mode plasma it may be possible to reduce the energy expelled per ELM, mitigating damage to the tokamak parts.One technique for mitigating these risks was first implemented successfully in TCV Ohmic H-mode plasmas, using the so called ELM pacing with vertical kicks [1]. TCV vertical stabilization coils located inside the vessel proved to be well suited to apply the necessary perturbations to induce vertical displacements of the plasma. This method was then reproduced on AUG [2,3] and later on JET [4] with significant improvements.During 2016 a set of experiments were planned to further study the physics of ELM pacing taking advantage of the new neutral beam heating (NBH) system installed on TCV [5]. This submission reports on the related preparation of the TCV digital vertical stabilization control system [6] actuating the in-vessel coils. A number of tools were deployed to drive a fast radial-field variation and preliminary experimental validation was carried out. The tools and simulations, as well as preliminary experiments aiming at evaluating the controller's accuracy are reported.

Multi-scale gyrokinetic simulations of an Alcator C-Mod, ELM-y H-mode plasma

High fidelity, multi-scale gyrokinetic simulations capable of capturing both ion () and electron-scale () turbulence were performed in the core of an Alcator C-Mod ELM-y H-mode discharge which exhibits reactor-relevant characteristics. These simulations, performed with all experimental inputs and realistic ion to electron mass ratio () provide insight into the physics fidelity that may be needed for accurate simulation of the core of fusion reactor discharges. Three multi-scale simulations and series of separate ion and electron-scale simulations performed using the GYRO code (Candy and Waltz 2003 J. Comput. Phys. 186 545) are presented. As with earlier multi-scale results in L-mode conditions (Howard et al 2016 Nucl. Fusion 56 014004), both ion and multi-scale simulations results are compared with experimentally inferred ion and electron heat fluxes, as well as the measured values of electron incremental thermal diffusivities—indicative of the experimental electron temperature profile stiffness. Consistent with the L-mode results, cross-scale coupling is found to play an important role in the simulation of these H-mode conditions. Extremely stiff ion-scale transport is observed in these high-performance conditions which is shown to likely play and important role in the reproduction of measurements of perturbative transport. These results provide important insight into the role of multi-scale plasma turbulence in the core of reactor-relevant plasmas and establish important constraints on the the fidelity of models needed for predictive simulations.

Dynamics and stability of divertor detachment in H-mode plasmas on JET

The dynamics and stability of divertor detachment in seeded, type-I, ELMy H-mode plasmas with dominant NBI heating in the JET ITER-like wall device is studied by means of an integrated analysis of diagnostic data from several systems, classifying data relative to the ELM times. It is thereby possible to study the response of the detachment evolution to the control parameters (SOL input power, upstream density and impurity fraction) prevailing during the inter-ELM periods and the effect of ELMs on the detached divertor. A relatively comprehensive overview is achieved, including the interaction with the targets at various stages of the ELM cycle, the role of ELMs in affecting the detachment process and the overall performance of the scenario. The results are consistent with previous studies in devices with an ITER-like, metal wall, with the important advance of distinguishing data from intra- and inter-ELM periods. Operation without significant degradation of the core confinement can be sustained in the presence of strong radiation from the x-point region (MARFE).

Favourable scenarios established by SMBI for the realization of the ELMy H-mode at HL-2A**Project supported by the National Natural Science Foundation of China (Grant Nos. 11375057 and 11175061) and the National Magnetic Confinement Fusion Science Program, China (Grant Nos. 2010GB102003 and 2014GB108003).

The ELMy H-mode plasmas realized with the supersonic molecular beam injection (SMBI) are studied in relation to the energy confinement and the heating power for the L–H transition ( in the HL-2A tokamak. A database is assembled for this study based on the ELMy H-mode discharges during the experimental campaigns in the period 2009–2013 at the HL-2A tokamak. The statistical results show that the SMBI is favourable for reaching the H-mode by reducing the heating power at the L–H transition and for the H-mode performance by improving the energy confinement compared with the ordinary gas puffing (GP). The reduction of is about 20% when the density is low, and the energy confinement enhancement factor of 1.5 is achieved with the SMBI. Note that in the database the density dependence of is non-monotonic with the at which the is minimum. Most of data are on the low density branch where the increases with the decrease in density. The minimum of the in HL-2A is comparable to the ITPA multi-machine threshold power scaling . The physics behind the reduction of the with the SMBI is also investigated in relation to the change of the density gradient at the plasma edge, the gas fuelling efficiency, and the recycling.

Modification of adhered dust on plasma-facing surfaces due to exposure to ELMy H-mode plasma in DIII-D

Transient heat load tests have been conducted in the lower divertor of DIII-D using DiMES manipulator in order to study the behavior of dust on tungsten Plasma Facing Components (PFCs) during ELMy H-mode discharges. Samples with pre-adhered, pre-characterized dust have been exposed at the outer strike point (OSP) in a series of discharges with varied intra-(inter-) ELM heat fluxes. We used C dust because of its high sublimation temperature and non-metal properties. Al dust as a surrogate for Be and W dust were employed as relevant to that in the ITER divertor. The poor initial thermal contact between the substrate and the particles led to overheating, sublimation and shrinking of the carbon dust, and wetting induced coagulation of Al dust. Little modification of the W dust was observed. An enhanced surface adhesion and improvement of the thermal contact of C and Al dust were the result of exposure. A post mortem “adhesive tape” sampling showed that 70% of Al, <5% of W and C particles could not be removed from the surface owing to the improved adhesion. Al and C but not W particles that could be lifted had W inclusions indicating damage to the substrate. This suggests that non destructive methods may be inefficient for removal of dust in ITER.

Observation of internal transport barrier in ELMy H-mode plasmas on the EAST tokamak

The internal transport barrier (ITB) has been obtained in ELMy H-mode plasmas by neutron beam injection and lower hybrid wave heating on the Experimental Advanced Superconducting Tokamak (EAST). The ITB structure has been observed in profiles of ion temperature, electron temperature, and electron density within ρ < 0.5. It was also observed that the ITB formation is stepwise. Due to the ITB formation, the confinement quality H98y2 increases from 1 to 1.1 and the normalized beta, βN, increases from 1.5 to near 2. The fishbone activity observed during the ITB phase suggests the central safety factor q(0) ∼ 1. Transport coefficients are calculated by particle balance and power balance analysis, showing an obvious reduction after the ITB formation.