Surface chemical/binding reaction of coated Li layer by lithium vapor injectors in LIGHT-1

The Lithium Injection Gettering of Hydrogen and its Transport (LIGHT-1) experiment has begun at the NIFS. To study the material probes installed in the cylindrical vacuum chamber, the chemical characteristics for lithium are analyzed using X-ray photoelectron spectroscopy (XPS). The characteristics of chemical binding between lithium and other impurities are shown to be oxide bindings. In addition, the influence of the vacuum vent effect due to exposure to air was determined in both solid lithium and lithium-coated probes in LIGHT-1. Using the peak positions of Li₂O and pure lithium, the thickness of the coated lithium is estimated. For the SS316 target, the coated lithium shows two different peaks, Li1s and Fe3p, located at a similar binding energy region. Thus, the real lithium intensities can be measured by the separation of the peaks. After this analysis, the coated thickness of lithium is estimated to be from 8 to 20 nm, and it is not uniform in the Z-axis direction, probably due to erosion by glow discharge.     

Lithium coatings on NSTX plasma facing components and its effects on boundary control, core plasma performance, and operation

NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.     

Thermal control of the liquid lithium divertor for NSTX

The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with 0.25 mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500 W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.     

Influence of lithium coatings with large-area coverage on EAST plasma performance

Over the last two EAST campaigns, lithium coatings by oven evaporation were carried out as a routine wall conditioning method and significant progresses has been achieved. By upgrading the EAST lithium coating systems, lithium area coverage increased from ∼35% in 2010 to ∼85% in 2012. Accompanying the increased lithium coverage, carbon, oxygen and molybdenum impurities were decreased to extremely low levels. In addition, hydrogen concentration was further decreased with the H/(H + D) ratio falling as low as 2.5%. The effective recycling coefficient decreased step-by-step to ∼0.89 and remained below unity for ∼100 discharges. This allowed for effective feedback control of the plasma density. The wall retention rate increased from 55% to 75%, which also indicated stronger pumping of deuterium particles with increased Li coverage. With the help of increased lithium coverage, H-mode plasmas were generally easier to obtain and the EAST parameter space was enlarged.     

Outgassing from alpha particle irradiation of lithium hydride and lithium hydroxide

We have experimentally studied the effects of α-particle radiation on isotopically enriched lithium hydride (⁶LiH) and its corrosion product lithium hydroxide (⁶LiOH) to determine, in particular, the type and amount of gases evolved during irradiation. SRIM Monte Carlo simulations suggest that irradiating these materials with 2.2-MeV α-particles will ionize atoms and form hydrogen vacancies in the target material, and that these α-particles will penetrate 13.5 and 9 μm into LiH and LiOH, respectively. Using an accelerator to irradiate LiH and LiOH with 2.2-MeV α-particles released only H₂ and CO₂; no other product gases were observed. At 25 °C, doses that simulated 66.5 years of actinide exposure (with accelerated fluxes) produced 2.3 × 10⁵ mol/(cm³ J) H₂ in LiH and 2.3 × 10⁶ mol/(cm³ J) H₂ in LiOH, in the form of a ∼9-μm-thick surface layer on LiH. More H₂ evolved from LiOH than from LiH. We argue that the production of H₂ gas was the result of radiolysis, rather than radiation-induced chemical reaction.     

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n_e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n_e/n_GW  1), to enable n_e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n_e/n_GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m²) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.     

Erosion of lithium coatings on TZM molybdenum and graphite during high-flux plasma bombardment

The rate at which Li films will erode under plasma bombardment in the NSTX-U divertor is currently unknown. It is important to characterize this erosion rate so that the coatings can be replenished before they are completely depleted. An empirical formula for the Li erosion rate as a function of deuterium ion flux, incident ion energy, and Li temperature was developed based on existing theoretical and experimental work. These predictions were tested on the Magnum-PSI linear plasma device capable of ion fluxes >10²⁴ m⁻² s⁻¹, ion energies of 20 eV and Li temperatures >800 °C. Li-coated graphite and TZM molybdenum samples were exposed to a series of plasma pulses during which neutral Li radiation was measured with a fast camera. The total Li erosion rate was inferred from measurements of Li-I emission. The measured erosion rates are significantly lower than the predictions of the empirical formula. Strong evidence of fast Li diffusion into graphite substrates was also observed.     

Development of technology for recovering lithium from seawater by electrodialysis using ionic liquid membrane

Tritium fuel for fusion reactors is produced by reacting lithium-6 (⁶Li) with neutrons in tritium breeders. This study demonstrates a method for Li recovery from seawater, wherein Li does not permeate from the anode side to the cathode side through an ionic liquid N,N,N-trimethyl-N-propylammonium–bis(trifluoromethanesulfonyl) imide. Almost all Li ions remain on the anode side (seawater), whereas the other ions in the seawater permeate to the cathode side through the ionic liquid with an applied electric voltage of 2–3 V.     

大功率NBI系统的PLC时序控制应用

论述了利用PLC逻辑关系对放电实验控制运行的工作原理,介绍了PLC梯形图应用程序和VB6.0环境下上位机监控程序的开发以及良好的人机操作界面,通过上位机监控界面来实时监控各电源及设备的运行状态.PLC控制系统的应用保障了实验装置的安全运行,极大地方便了实验中对放电参数的改变和设置.     

A review of plasma-wall boundary effects on core confinement and lithium applications to boundary-controlled magnetic fusion experiments

Reduced particle recycling in the edge region is often considered to be a necessary condition to achieve high-performance confinement in magnetic fusion experiments. However, it is also true that whenever or however core confinement improves, as a result edge recycling is reduced. To provide a logical interpretation of this circular cause-and-consequence relation, reviewed in this paper are some of the important data demonstrating plasma-wall boundary effects on core confinement and also those evaluating innovative wall concepts with lithium applied to sustain optimized boundary conditions for the operation of steady-state fusion power reactors.     

Design of flowing liquid lithium device for HT-7 tokamak

Lithium is a very attractive element due to its very low radiation power, strong H retention as well as strong O getter activity. Flowing liquid lithium (FLiLi) device, to be used as a plasma-facing limiters, has been designed and will be tested in HT-7 tokamak. It is mainly composed of distributor, guide plate, collector, and heater as well as cooling loop. The heater uses heater strip and cooling loop design, to control the temperature of lithium on the guide plate ranging from 200 °C to 400 °C. The distributor attached to feeding pipe, distributes liquid lithium (LiLi) flowing on the guide plate. The collector was designed to reclaim the superfluous LiLi and transport it out of device.The paper focuses on the design of flowing liquid lithium device. In addition to the process of design, thermal analysis has been carried out using finite element method (FEM) for optimizing the structure of heater and cooling loop and results of analysis are presented.     

Implications of NSTX lithium results for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last five years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a lithium evaporation system which can evaporate up to ∼100 g of lithium onto the lower divertor plates between lithium re-loadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, edge localized mode (ELM) control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.     

微波烧结偏铝酸锂陶瓷的可行性研究

偏铝酸锂(γ-LiAlO2)陶瓷，具有良好的热稳定性、化学稳定性、辐照稳定性和机械强度，以及与其他材料的广泛相容性。采用了微波烧结γ-LiAlO2陶瓷，避免了试件出现裂纹与变形，实现了快速烧结，并得到了传统电炉烧结无法比拟的优异的力学和物理性能，如抗压强度、气孔率以及晶粒尺寸。这为以后高效高质生产偏铝酸锂陶瓷提供了新的途径。     

Functions of intrinsic magnetic field line structures on lithium ion transport in the LHD plasma periphery

Plasma discharge operation with lithium coating suggests that the lithium effectively control neutral particles in the plasma periphery, which can lead to improvement of plasma parameters. The effect of lithium coating on the large helical device (LHD) for a closed helical divertor configuration is discussed from viewpoints of neutral particle and impurity ion transport in the plasma periphery. It shows that the closed helical divertor configuration can enhance the neutral particle density in the divertor region, which is enough to achieve efficient particle control, and that it can effectively confine neutral lithium atoms near divertor plates. A one-dimensional impurity (lithium) ion transport analysis along magnetic field lines on divertor legs indicates that the friction force due to the plasma flow from the main plasma is dominant over the thermal force caused by the temperature gradient on the divertor legs, which prevents lithium ion contamination in the main plasma and excessive cooling of the plasma temperature in an ergodic layer. The analysis shows that the lithium coating is compatible with LHD plasma discharge operation for the closed helical divertor configuration.     

Progress in development and application of lithium based components for Tokamak

Experiments with lithium plasma facing components (PFCs) show promising results for the operation of hot plasma facilities and the general improvement of plasma parameters. The design and development of new tokamak plasma facing material (PFM) based on lithium capillary porous systems (CPS) are described in this paper.The recent progress in the development of limiters with different kinds of CPS is relevant for protection of tokamak PFCs from damage under normal operation, ELMs and disruptions. New PFM eliminates the lithium flux into plasma, its pollution and lithium accumulation.Here we present an overview of the design and the experimental tests of the liquid lithium limiters. These limiters are based on CPS with hard matrix from stainless steel mesh, molybdenum and tungsten. Different types of limiter have been taken into account: the horizontal and vertical rail type limiters with passive and active cooling for investigation the possibility to provide the closed lithium circulation in tokamak chamber; the ring CPS-based limiter for investigation of lithium behavior in limiter scrape-off layer (SOL).Here we also present the preliminary results of the application of the cryogenic techniques for lithium removal from the chamber wall after operation in hot plasma.     

Investigation of lithium as plasma facing materials on HT-7

First experiment of liquid lithium limiter was successfully carried out on HT-7 tokamak and a few positive results were obtained. The results showed that by using lithium limiter, specially liquid lithium limiter, Hα intensity reduced 20-30%, the emission of CIII and OV decreased about 10-20%, loop voltage had a slight decline, the core electron temperature slightly increased, the particle confinement time increased by a factor of 2, and the energy confinement time increased 20%. After lithium coating, the hydrogen recycling decreased, and core electron temperature increased significantly by a factor of 2. At the same time, after lithium coating, electron density of edge plasmas obviously decreased while electron temperature slightly increased. These encouraging results are very useful for further research of long tray lithium limiter on HT-7 and liquid divertor on EAST.     

Peaked density profiles and MHD activity on FTU in lithium dominated discharges

Density profiles become broader, as the line averaged density is increased. At higher density, change of the trend is associated with the appearance of the MARFE. Lithium coated wall extends the maximum density accessible at low current (well above the Greenwald density limit) and can produce profiles with very high density gradient. At higher current the effect on the density limit can be exceeded only if the magnetic field is raised too. A comparison of successive similar discharges before and after wall conditioning with lithium showed a reduction of the MHD activity.     

Experience and technical issues of liquid lithium application as plasma facing material in tokamaks

The following critical issues of liquid lithium used in tokamak conditions are considered: major physical properties of lithium, physico-chemical aspects of lithium interaction and compatibility with structural materials of fusion reactors. Lithium capillary-porous system (CPS) is considered as advanced plasma facing material for power fusion reactor and its main properties are presented. Review of plasma facing element (PFE) structures based on lithium CPS and tests results in T-11M, T-10 and FTU tokamaks are included. Brief review of projects of lithium limiter of FTU with active system for thermal stabilization and module of lithium divertor for KTM tokamak with liquid metal (Na-K) cooling system based on the lithium CPS use are presented.     

Wall conditioning strategies in the stellarator TJ-II

This paper summarizes the different wall conditioning strategies applied in the TJ-II stellarator since the operational starting in 1997. In a first stage (1997-2001) the all-metal machine (stainless steel) was conditioned by He glow discharge. This procedure allowed an acceptable density control in low power ECRH plasmas. The boronization of TJ-II (starting in 2001) led to obtain low Z plasmas with high ECRH injected power, but the density control under NBI injection was not possible. Finally, the coating of the boronized walls with a lithium thin film has allowed to obtain high reproducible NBI plasmas with good density control.     

痕量锂同位素的热表面电离质谱高精度测量

建立了痕量锂同位素的高精度热电离质谱测量技术。通过双带测量、加入磷酸发射剂及采用预烧处理方法等途径,抑制了分馏效应,提高了痕量锂质谱分析的精度。采用浓缩锂同位素标准样品考察了测量效果,对于100ng锂样品,测量相对标准偏差好于0.086%;对于10ng锂样品,相对标准偏差好于0.90%。