R. Ding

Characterizations of power loads on divertor targets for type-I, compound and small ELMs in the EAST superconducting tokamak

The Experimental Advanced Superconducting Tokamak (EAST) has recently achieved a variety of H-mode regimes with different edge-localized mode (ELM) dynamics, including type-I ELMs, compound ELMs, which are manifested by the onset of a large spike followed by a sequence of small spikes on Dα emissions, usual type-III ELMs, and very small ELMs. This newly observed very small ELMy H-mode appears to be similar to the type-II ELMy H-mode, with higher repetition frequency (~1 kHz) and lower amplitude than the type-III ELMy H-mode, exhibiting an intermediate confinement level between type-I and type-III ELMy H-modes. The energy loss and divertor power load are systematically characterized for these different ELMy H-modes to provide a physics basis for the next-step high-power long-pulse operations in EAST. Both type-I and compound ELMs exhibit good confinement (H98(y,2) ~ 1). A significant loss of the plasma stored energy occurs at the onset of type-I ELMs (~8%) and compound ELMs (~5%), while no noticeable change in the plasma stored energy is observed for the small ELMs, including both type-III ELMs and very small ELMs. The peak heat flux on divertor targets for type-I ELMs currently achieved in EAST is about 10 MW m−2, as determined from the divertor-embedded triple Langmuir probe system with high time resolution. As expected, type-III ELMs lead to much smaller divertor power loads with a peak heat flux of about 2 MW m−2. Peak power loads for compound ELMs are between those for type-I and type-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MW m−2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may promote particle and power exhaust throughout the very small ELMy H-mode regime.

Material migration studies with an ITER first wall panel proxy on EAST

The ITER beryllium (Be) first wall (FW) panels are shaped to protect leading edges between neighbouring panels arising from assembly tolerances. This departure from a perfectly cylindrical surface automatically leads to magnetically shadowed regions where eroded Be can be re-deposited, together with co-deposition of tritium fuel. To provide a benchmark for a series of erosion/re-deposition simulation studies performed for the ITER FW panels, dedicated experiments have been performed on the EAST tokamak using a specially designed, instrumented test limiter acting as a proxy for the FW panel geometry. Carbon coated molybdenum plates forming the limiter front surface were exposed to the outer midplane boundary plasma of helium discharges using the new Material and Plasma Evaluation System (MAPES). Net erosion and deposition patterns are estimated using ion beam analysis to measure the carbon layer thickness variation across the surface after exposure. The highest erosion of about 0.8??m is found near the midplane, where the surface is closest to the plasma separatrix. No net deposition above the measurement detection limit was found on the proxy wall element, even in shadowed regions. The measured 2D surface erosion distribution has been modelled with the 3D Monte Carlo code ERO, using the local plasma parameter measurements together with a diffusive transport assumption. Excellent agreement between the experimentally observed net erosion and the modelled erosion profile has been obtained.

Cleaning of HT-7 Tokamak Exposed First Mirrors by Radio Frequency Magnetron Sputtering Plasma

The stainless steel (SS) first mirror pre-exposed in the deposition-dominated environment of the HT-7 tokamak was cleaned in the newly built radio frequency (RF) magnetron sputtering plasma device. The deposition layer on the FM surface formed during the exposure was successfully removed by argon plasma with a RF power of about 80 W and a gas pressure of 0.087 Pa for 30 min. The total reflectivity of the mirrors was recovered up to 90% in the wavelength range of 300–800 nm, while the diffuse reflectivity showed a little increase, which was attributed to the increase of surface roughness in sputtering, and residual contaminants. The FMs made from single crystal materials could help to achieve a desired recovery of specular reflectivity in the future.

Influence of helium puff on divertor asymmetry in Experimental Advanced Superconducting Tokamak

Divertor asymmetries with helium puffing are investigated in various divertor configurations on Experimental Advanced Superconducting Tokamak (EAST). The outer divertor electron temperature decreases significantly during the gas injection at the outer midplane. As soon as the gas is injected into the edge plasma, the power deposition drops sharply at the lower outer target while increases gradually at the lower inner target in LSN configuration; the power deposition increases quickly at the upper outer target while remains unchanged at the upper inner target in upper single null configuration; the power deposition increases slightly at the outer targets while shows no obvious variation at the inner targets in double null configuration. The radiated power measured by the extreme ultraviolet arrays increases significantly due to helium gas injection, especially in the outer divertor. The edge parameters are measured by reciprocating probes at the outer midplane, showing that the electron temperature and density increase but the parallel Mach number decreases significantly due to the gas injection. Effects of poloidal E × B drifts and parallel SOL flows on the divertor asymmetry observed in EAST are also discussed.

Thermal analysis on the EAST tungsten plasma facing components with shaping structure counteracting the misalignment issues

Tungsten monoblock type tiles with ITER dimensions along with supporting cassette components were installed at EASTs upper diverter during 2014 and EASTs lower diverter will also be upgraded in the future. These cassette structures pose critical issues on the high cumulative incident heat flux due to the leading edges and misalignments (0 ~ 1.5 mm), which may result in the destruction or even melting of the tungsten tile. The present work summarizes the thermal analysis using ANSYS multiphysics software 15.0 performed on the actively cooled W tiles to evaluate the shaping effect on surface temperature. In the current heat flux conditions (Q|| ~ 100 MW m−2), the adopted chamfer shaping (1 × 1 mm) can only reduce the maximum temperature by about 14%, but it also has a melting risk at the maximum misalignment of 1.5 mm. The candidate shaping solutions elliptical (round) edge, dome and fish-scale are analyzed for comparison and are identified not as good as the dual chamfer structure. A relatively good dual chamfer (2 × 13 mm) shaping forming a symmetrical sloping roof structure can effectively counteract the 1.5 mm misalignment, reducing the maximum temperature by up to 50%. However, in the future heat flux conditions (Q|| ~ 287 MW m−2), it may only endure about 0.5 mm misalignment. Moreover, no proper shaping solution has been found that can avoid melting at the maximum misalignment of 1.5 mm. Thus, the engineering misalignment has to be limited to an acceptable level.

PIC-EDDY Simulation of Different Impurities Deposition in Gaps of Carbon Tiles

A 3D Monte Carlo (MC) code PIC- EDDY, based on EDDY (erosion and deposition dynamic simulation) code, was used to investigate the redeposition of different impurities in the gaps of C tiles. By incorporating the rate coefficients of beryllium (Be) and tungsten (W) into the code, we obtain deposition profiles of hydrocarbon, beryllium and tungsten particles in the toroidal and poloidal gaps, respectively. The redeposition rate of tungsten was found to be higher than those of other impurities in the gaps, except at the bottom, due to its easier local deposition within one gyroradius. Due to the effect of reflection coefficients of hydrocarbon fragments on graphite, fewer hydrocarbons were resided at the entrance while more were deposited on the sides of the gap. At elevated plasma temperatures (such as 30 eV), asymmetric deposition distributions were observed between the toroidal and poloidal gaps due to the dominant ionized particles. Ions were mainly deposited within 1 mm depth inside gaps, and the bottom deposition particles were almost all neutrals.

Development of a time-of-flight low-energy neutral particle analyzer for EAST tokamak

To understand the erosion effect of neutral particles on the first wall, a low energy neutral particle analyzer (LENPA), based on the time-of-flight method, has been developed for the Experimental Advanced Superconducting Tokamak (EAST). The LENPA to be installed in the EAST mid-plane mainly consists of a chopper sub-system, a 3 m long flight tube, two sets of detector assemblies, and data acquisition and processing, vacuum, power supply, and control sub-systems. The neutral outflux is gated in bunches of 1 μs time scale by a slotted rotating disc which is driven by a vacuum compatible motor modified from a turbomolecular pump. A He-Ne laser beam is projected through the disc slit to record the instants of chopper slits opening with an avalanche photodiode module. An on-axis electron multiplier detects chopped neutrals, and a central perforated Cu-Be plate is employed to channel the emitted secondary electrons into an off-axis electron multiplier. The radiation peaks of on-axis electron multiplier caused by UV photons projected through the central hole of the Cu-Be plate provide an alternative way to record the chopper slits opening time. With an additional 4 TB fast memory card, 1 GS/s sampling rate has been realized by using a GaGe acquisition card continuously. The LENPA data from the EAST together with neutral particle material erosion experiments will be used to benchmark the simulation results for better predictions on future fusion reactors, such as ITER and China Fusion Engineering Test Reactor (CFETR).

ERO modelling of tungsten erosion and re-deposition in EAST L mode discharges

Tungsten erosion and re-deposition at the upper outer divertor of the Experimental Advanced Superconducting Tokamak has been modelled using the 3D Monte Carlo code ERO. The measured divertor plasma condition in attached L mode discharges with upper single null configuration has been used to build the background plasma in the simulations. The tungsten gross erosion rate is mainly determined by carbon impurity in the background plasma. Increasing carbon concentration can first increase and afterwards suppress the tungsten erosion rate. Taking into account the material mixing surface model, the influence of eroded particles returning to the surface on sputtering has been studied. Sputtering by eroded particles returning to the surface can significantly enhance the gross erosion by reduction of the carbon ratio within the surface interaction layer and by increasing the erosion rate due to sputtering by both eroded tungsten and carbon particles. Modelling indicates that carbon deposition occurs on the dome pla...

Modeling of advanced divertor configuration on experimental advanced superconducting tokamak by SOLPS5.0/B2.5-Eirene

Heat exhaust is one of the most challenging issues to be addressed for tokamak magnetic confinement fusion research. Detailed modeling with SOLPS5.0/B2.5-Eirene code package is carried out to examine an alternative advanced divertor configuration, i.e., quasi snowflake (QSF), for long pulse operation in EAST. Comparison is also made with the lower single null (LSN) divertor configuration. SOLPS predicts that the quasi snowflake configuration significantly reduces the peak heat flux at the lower divertor outer target, by a factor of 2–3, owing to the magnetic flux expansion. Furthermore, the density threshold for detachment is much lower for QSF, compared to LSN under the same upstream conditions. This indicates that QSF provides a promising tool for controlling heat flux at divertor target while maintaining a lower separatrix density, which is highly desirable for current drive, thus greatly facilitating long-pulse operation in EAST.

Influences of Dispersed Lanthanum Oxide Additive on the Properties of Tungsten-Based Plasma-Facing Material

Abstract To investigate the influences of dispersed lanthanum oxide (La2O3) additive on the properties of a tungsten (W)-based plasma-facing material, ultrafine-grained W-1% La2O3 composite has been successfully fabricated using the resistance sintering under ultrahigh pressure method, which can suppress W grain growth during sintering processes. Its relative density, Vickers microhardness, microstructure, and thermal conductivity have been analyzed and compared with those of pure W. Moreover, its behaviors under fusion-related conditions, i.e., edge plasma loading in the HT-7 tokamak and transient heat flux simulated by a high-intensity pulsed ion beam, have been evaluated. It is shown that without the fine-grain strengthening effect of dispersed particles, the La2O3 additive as second-phase particles being dispersed in W-based plasma-facing material degrades the material resistance ability under plasma heat loading.