J.S. Hu

First observations of ELM triggering by injected lithium granules in EAST

The first results of edge-localized mode (ELM) pacing using small spherical lithium granules injected mechanically into H-mode discharges are reported. Triggering of ELMs was accomplished using a simple rotating impeller to inject sub-millimetre size granules at speeds of a few tens of meters per second into the outer midplane of the EAST fusion device. During the injection phase, ELMs were triggered with near 100% efficiency and the amplitude of the induced ELMs as measured by Dα was clearly reduced compared to contemporaneous naturally occurring ELMs. In addition, a wide range of granule penetration depths was observed. Moreover, a substantial fraction of the injected granules appeared to penetrate up to 50% deeper than the 3 cm nominal EAST H-mode pedestal width. The observed granule penetration was, however, less deep than suggested by ablation modelling carried out after the experiment. The observation that ELMs can be triggered using the injection of something other than frozen hydrogenic pellets allows for the contemplation of lithium or beryllium-based ELM pace-making on future fusion devices. This change in triggering paradigm would allow for the decoupling of the ELM-triggering process from the plasma-fuelling process which is currently a limitation on the performance of hydrogen-based ELM mitigation by injected pellets.

Approaches towards long-pulse divertor operations on EAST by active control of plasma–wall interactions

The Experimental Advanced Superconducting Tokamak (EAST) has demonstrated, for the first time, long-pulse divertor plasmas over 400 s, entirely driven by lower hybrid current drive (LHCD), and further extended high-confinement plasmas, i.e. H-modes, over 30 s with predominantly LHCD and advanced lithium wall conditioning. Many new and exciting physics results have been obtained in the quest for long-pulse operations. The key findings are as follows: (1) access to H-modes in EAST favours the divertor configuration with the ion ∇B drift directed away from the dominant X-point; (2) divertor asymmetry during edge-localized modes (ELMs) also appears to be dependent on the toroidal field direction, with preferential particle flow opposite to the ion ∇B drift; (3) LHCD induces a striated heat flux (SHF), enhancing heat deposition away from the strike point, and the degree of SHF can be modified by supersonic molecule beam injection; (4) the long-pulse H-modes in EAST exhibit a confinement quality between type-I and type-III ELMy H-modes, with H98(y,2) ~ 0.9, similar to type-II ELMy H-modes.

First results of the use of a continuously flowing lithium limiter in high performance discharges in the EAST device

As an alternative choice of solid plasma facing components (PFCs), flowing liquid lithium can serve as a limiter or divertor PFC and offers a self-healing surface with acceptable heat removal and good impurity control. Such a system could improve plasma performance, and therefore be attractive for future fusion devices. Recently, a continuously flowing liquid lithium (FLiLi) limiter has been successfully designed and tested in the EAST superconducting tokamak. A circulating lithium layer with a thickness of <0.1 mm and a flow rate ~2 cm3 s−1 was achieved. A novel in-vessel electro-magnetic pump, working with the toroidal magnetic field of the EAST device, was reliable to control the lithium flow speed. The flowing liquid limiter was found to be fully compatible with various plasma scenarios, including high confinement mode plasmas heated by lower hybrid waves or by neutral beam injection. It was also found that the controllable lithium emission from the limiter was beneficial for the reduction of recycling and impurities, for the reduction of divertor heat flux, and in certain cases, for the improvement of plasma stored energy, which bodes well application for the use of flowing liquid lithium PFCs in future fusion devices.

First results of flowing liquid lithium limiter in HT-7

Two different types of flowing liquid lithium limiters were firstly installed and successfully tested in HT-7 tokamaks in 2012 and some encouraging results were obtained. Two limiters of the first type, called FLiLi limiters, used a thin lithium layer flowing under gravity. The other type had lithium-metal infused trenches (LIMIT) for thermoelectric magnetohydrodynamic drive of the liquid metal flow. The surface of one of the FLiLi limiters was coated by evaporated lithium before liquid lithium was injected by Ar pressure into a special distributor of the limiter. Then the liquid lithium could slowly move along the plasma facing guide surface of the limiter due to gravity. For LIMIT, it was found that liquid lithium could flow along the trenches as expected with a velocity of about 3.7 ± 0.5 cm s−1 driven by the electromagnetic force, which came from the interaction between the thermoelectric current and magnetic field. Use of flowing liquid lithium limiters in HT-7 resulted in reduction of particle recycling, suppression of impurity emission and improvement of the confinement.

Recent advances in long-pulse high-confinement plasma operations in Experimental Advanced Superconducting Tokamaka)

A long-pulse high confinement plasma regime known as H-mode is achieved in the Experimental Advanced Superconducting Tokamak (EAST) with a record duration over 30u2009s, sustained by Lower Hybrid wave Current Drive (LHCD) with advanced lithium wall conditioning and divertor pumping. This long-pulse H-mode plasma regime is characterized by the co-existence of a small Magneto-Hydrodynamic (MHD) instability, i.e., Edge Localized Modes (ELMs) and a continuous quasi-coherent MHD mode at the edge. We find that LHCD provides an intrinsic boundary control for ELMs, leading to a dramatic reduction in the transient power load on the vessel wall, compared to the standard Type I ELMs. LHCD also induces edge plasma ergodization, broadening heat deposition footprints, and the heat transport caused by ergodization can be actively controlled by regulating edge plasma conditions, thus providing a new means for stationary heat flux control. In addition, advanced tokamak scenarios have been newly developed for high-performance ...

Scaling of divertor power footprint width in RF-heated type-III ELMy H-mode on the EAST superconducting tokamak

Dedicated experiments for the scaling of divertor power footprint width have been performed in the ITER-relevant radiofrequency (RF)-heated H-mode scheme under the lower single null, double null and upper single null divertor configurations in the Experimental Advanced Superconducting Tokamak (EAST) under lithium wall coating conditioning. A strong inverse scaling of the edge localized mode (ELM)-averaged power fall-off width with the plasma current (equivalently the poloidal field) has been demonstrated for the attached type-III ELMy H-mode as λq ∝ I −1.05 p by various heat flux diagnostics including the divertor Langmuir probes (LPs), infra-red (IR) thermograph and reciprocating LPs on the low-field side. The IR camera and divertor LP measurements show that λq,IR ≈ λq,div-LPs/1.3 = 1.15B −1.25 p,omp , in good agreement with the multi-machine scaling trend during the inter-ELM phase between type-I ELMs or ELM-free enhanced Dα (EDA). H-mode. However, the magnitude is nearly doubled, which may be attributed to the different operation scenarios or heating schemes in EAST, i.e., dominated by electron heating. It is also shown that the type-III ELMs only broaden the power fall-off width slightly, and the ELM-averaged width is representative for the inter-ELM period. Furthermore, the inverse Ip (Bp) scaling appears to be independent of the divertor configurations in EAST. The divertor power footprint integral width, fall-off width and dissipation width derived from EAST IR camera measurements follow the relation, λint ∼ λq +1.64S, yielding λ EAST = (1.39±0.03)λ EAST +(0.97±0.35) mm. Detailed analysis of these three characteristic widths was carried out to shed more light on their extrapolation to ITER.

RF wall conditioning – a new technique for future large superconducting tokamak

A new technique for wall conditioning with a high toroidal magnetic field has been developed on HT-7 superconducting tokamak by launching ion cyclotron resonant wave (ICRF) into plasmas. The RF wall conditioning techniques include hydrogen removal, isotopic and impurity control, boronization and siliconization, which have been proved to be very effective and could be routinely used in daily experiments.

On-surface manipulation of atom substitution between cobalt phthalocyanine and the Cu(111) substrate

On-surface fabrication of controllable nanostructures is an appealing topic in the field of molecular electronics. Herein, the adsorption of cobalt phthalocyanine (CoPc) on a Cu(111) surface is investigated utilizing a combination of photoelectron spectroscopy (PES) and density functional theory (DFT). Interestingly, the scenario of atom exchange is discovered at the interface at room temperature (RT), namely the substitution of the cobalt atom in CoPc by a surface Cu adatom. Moreover, thermal annealing enhances the substitution process considerably which is demonstrated to be complete at about 573 K. As revealed by DFT calculations, the driving force for the observed interface transmetalation is most probably provided by the initial strong molecular-substrate interaction between Co atoms and the Cu(111) surface, the external thermodynamic energy gained from thermal sublimation and thermal annealing, and the tendency to form Co–Cu alloy at the interface. While CoPc has been successfully utilized in electrocatalysts for fuel cell applications and CuPc is commonly used as a leading material in organic solar cells, this report of interface transmetalation from CoPc to CuPc in a solid state environment may offer an encouraging approach towards the artificial engineering of organometallic nanostructures and related properties for surface catalysts, molecular electronics and so on.

Conference report on the 3rd international symposium on lithium application for fusion devices

The third International Symposium on Lithium Application for Fusion Device (ISLA-2013) was held on 9–11 October 2013 at ENEA Frascati Centre with growing participation and interest from the community working on more general aspect of liquid metal research for fusion energy development. ISLA-2013 has been confirmed to be the largest and the most important meeting dedicated to liquid metal application for the magnetic fusion research. Overall, 45 presentation plus 5 posters were given, representing 28 institutions from 11 countries. The latest experimental results from nine magnetic fusion devices were presented in 16 presentations from NSTX (PPPL, USA), FTU (ENEA, Italy), T-11M (Trinity, RF), T-10 (Kurchatov Institute, RF), TJ-II (CIEMAT, Spain), EAST(ASIPP, China), HT-7 (ASIPP, China), RFX (Padova, Italy), KTM (NNC RK, Kazakhstan). Sessions were devoted to the following: (I) lithium in magnetic confinement experiments (facility overviews), (II) lithium in magnetic confinement experiments (topical issues), (III) special session on liquid lithium technology, (IV) lithium laboratory test stands, (V) Lithium theory/modelling/comments, (VI) innovative lithium applications and (VII) special Session on lithium-safety and lithium handling. There was a wide participation from the fusion technology communities, including IFMIF and TBM communities providing productive exchange with the physics oriented magnetic confinement liquid metal research groups. This international workshop will continue on a biennial basis (alternating with the Plasma–Surface Interactions (PSI) Conference) and the next workshop will be held at CIEMAT, Madrid, Spain, in 2015.

Erratum: New Steady-State Quiescent High-Confinement Plasma in an Experimental Advanced Superconducting Tokamak [Phys. Rev. Lett. 114, 055001 (2015)].

In our recently published work, there is an error in the stated direction of the edge coherent MHDmode (ECM) associated with real-time Li injection. This error does not affect the main results of the paper. The direction of the ECMwas obtained from two high sampling-frequencyMirnov magnetic pickup probes. In the Letter, we reported that the ECM propagates in the ion diamagnetic drift direction. The ECM actually propagates in the electron diamagnetic drift direction, consistent with the identified direction in our previous study [1].