Preliminary design and performance study of EAST liquid lithium limiter based on CPS

Lithium has the ability of H recycling suppression and impurities absorption and it can be used as plasma facing material (PFM) in tokamaks. Lithium conditioning experiments were launched on EAST, HT-7 and some other tokamaks for many years by using the methods of GDC, IRCF and evaporation. Liquid lithium has better performances in effective lifetime and heat removal aspects compared to non-liquid lithium. While, applying liquid lithium in the tokamak would cause the safety problem as the lithium can react with many substances violently and the magnetohydrodynamic behavior is difficult to be handled. EAST liquid lithium limiter (LLL) system is under developing and will be applied in EAST to study the main technologies of the liquid lithium application. The normal operation temperature of the limiter is expected as 230–550 °C under the active cooling of water. Capillary porous system (CPS) is used to prevent the lithium from splashing under large electromagnetic force by increasing the surface tension of the lithium. In order to investigate the cooling performance of the cooling design, the thermal-hydraulic analysis was done which shows that with 3 m/s flowing velocity, the water can keep the limiter under 550 °C all the time if the heat flux is lower than 0.7 MW/m². Under heat flux of 1 MW/m², the limiter should be retreated within 7 s to avoid erosion. The pressure drop of the coolant under 3 m/s is less than 40 kPa with temperature difference nearly 34 °C which meet the design requirements very well. The key manufacture process and technologies like vacuum bonding between the CuCrZr heat sink and 316L guide plate were well studied in the R&D process. The heating test on the test bench showed that the CPS can be heated efficiently by the heaters attached into the heat sink.     

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.     

A method to produce lithium pellets for fueling and ELM pacing in NSTX-U

A device for producing small, high frequency spherical droplets or pellets for lithium or other liquid metals has been developed and could aid in the controlled excitation or pacing of edge-localized plasma modes (ELMs). The Liquid Lithium/metal Pellet Injector (LLPI) could also be used to replenish lithium coatings of plasma-facing components (PFCs) during a plasma discharge. With NSTX-U having longer pulse lengths (up to 5 s), it is desirable to be able to inject lithium during the discharge to maintain the beneficial effects. Using a nozzle injector design and a surrogate to lithium, Wood's metal, the LLPI has achieved droplet diameters between 0.6 mm < d_drop < 1 mm in diameter and frequencies up to 1.5 kHz with argon gas driving the formation. This paper presents the LLPI being developed with initial results mainly using Wood's metal and some lithium, using high pressure argon to force the liquid lithium through the nozzle.     

Module of lithium divertor for KTM tokamak

Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium–metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z_eff of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) – new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of steady-state operating lithium divertor module project for Kazakhstan tokamak KTM. At present the lithium divertor module for KTM tokamak is under development in the framework of ISTC project # K-1561. Initial heating up to 200 °C and lithium surface temperature stabilization during plasma interaction in the range of 350–550 °C will be provided by external system for thermal stabilization due to circulation of the Na–K heat transfer media. Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Development, creation and experimental research of lithium divertor model for KTM will allow to solve existing problems and to fulfill the basic approaches to designing of lithium divertor and in-vessel elements of new fusion reactor generation, to investigate plasma physics aspects of lithium influence, to develop technology of work with lithium in tokamak conditions. Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation.     

Overview of recent developments in pellet injection for ITER

Pellet injection is the primary fueling technique planned for core fueling of ITER burning plasmas. Also, the injection of relatively small pellets to purposely trigger rapid small edge localized modes (ELMs) has been proposed as a possible solution to the heat flux damage from larger natural ELMs likely to be an issue on the ITER divertor surfaces. The ITER pellet injection system is designed to inject pellets into the plasma through both inner and outer wall guide tubes. The inner wall guide tubes will provide high throughput pellet fueling while the outer wall guide tubes will be used primarily to trigger ELMs at a high frequency (>15 Hz). The pellet fueling rate of each injector is to be up to 120 Pa m³/s, which will require the formation of solid D–T at a volumetric rate of ～1500 mm³/s. Two injectors are to be provided for ITER at the startup with a provision for up to six injectors during the D–T phase. The required throughput of each injector is greater than that of any injector built to date, and a novel twin-screw continuous extrusion system is being developed to meet the challenging design parameters. Status of the development activities is presented, highlighting recent progress.     

A repetitive pellet injection system for steady state fuelling in EAST superconducting tokamak

A new hydrogen/deuterium pellet injector has been developed for Experimental Advanced Superconducting Tokamak (EAST). The pellet injector based on a screw extruder is able to fire pellets (∅2 mm × 2 mm; frequency 1–10 Hz and velocity 150–300 m/s) in steady state mode with reliability greater than 95%. An injection line was designed for pumping propellant gas and for diagnostic purpose also. A guide tube for magnetic high-field side (HFS) injection was developed and theoretical calculation has been done. After successful engineering commissioning, the injection system served at EAST 2012 campaign and first experimental results were obtained.     

Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance

Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ～850 °C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5 cm of the lower vertical/horizontal divertor corner appear to have unused LiO bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500 nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10 nm.     

Li collection experiments on T-11M and T-10 in framework of Li closed loop concept

The concept of a steady state tokamak with plasma facing components (PFC) on the basis of liquid lithium circulation demands the decision of three tasks: lithium injection to the plasma, lithium ions collection before their deposition on the vacuum vessel and lithium returning to the injection zone. Main subject of paper is the investigations of Li collection by different types of limiters intersected the scrape-of-layer (SOL) in T-10 and T-11M tokamaks. For finding solution for this problem in T-11M and T-10, experiments have been applied with Li-, C-rail limiters and ring SS R-limiter-collector (T-11M). The efficiency of Li collection by limiters in T-11M and T-10 tokamaks was investigated by post mortem sample–witness analysis and (T-11M) by the use of the mobile graphite probe (limiter) as a recombination target in the stream of lithium ions. The characteristic depth of lithium penetration in the SOL area of T-11M is about 2 cm and 4 cm in SOL of T-10. The quantitative analysis of the sample–witnesses located on T-11M limiters showed that 60 ± 20% of the lithium injected during plasma operating of T-11M had been collected by limiters. It confirms an opportunity of the lithium ions collection by limiters in tokamak SOL.     

Lithium technologies for edge plasma control

We have investigated two new modes of operation been in T-10 limiter tokamak experiments with a novel rotary feeder of lithium dust. Quasi steady-state mode I and pulse mode II of dust delivery were realized in both OH and OH + ECRH disruption free plasmas at the lithium flow rate up to 2 × 10²¹ atoms/s. A higher flow rate in mode II with injection rate of ～5 × 10²¹ atoms/s caused a series of minor disruptions, which was completed by discharge termination after the major disruption. The observed decreases of bolometer and D_β signals, with increase of the electron density during the lithium dust injection, reveal the effects of the first wall conditioning. The lithium technology may provide inherent safety pathway for major disruption mitigation in a tokamak reactor, which requires demonstration in contemporary tokamak experiments.     

A distributed timing and synchronization system for EAST

The distributed timing and synchronization system (DTSS) plays an important role in Experimental Advanced Superconducting Tokamak (EAST), which is one of the national key fusion research facilities in China. This system synchronizes each subsystem of EAST by using reference clock and trigger. A prototype DTSS module has been developed based on PXI bus and RIO (reconfigurable I/O) devices. The DTSS can provide reference clock in frequency up to 80 MHz. The trigger can be pre-defined from 1 ms to 6872 s with 10 ns accuracy. In addition, this system can acquire, process signals, and send output or command to other systems. The DTSS has been successfully applied to 2010 fall EAST experiment, and the results confirmed its accuracy and reliability. After the analysis of system requirement, the architecture of the DTSS and the technical implementation based on PXI are presented in this paper.     

EAST ion cyclotron resonance heating system for long pulse operation

Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). The ICRF system for EAST has been developed to support long-pulse high-β advanced tokamak fusion physics experiments. The ICRF system is capable of delivering 12 MW 1000-s RF power to the plasma through two antennas. The phasing between current straps of the antennas can be adjusted to optimize the RF power spectrum. The main technical features of the ICRF system are described. Each of the 8 ICRF transmitters has been successfully tested to 1.5 MW for a wide range of frequency (25–70 MHz) on a dummy load. Part of the ICRF system was in operation during the EAST 2012 spring experimental campaign and a maximum power of 800 kW (at 27 MHz) lasting for 30 s has been coupled for long pulse H mode operation.     

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.     

Heat loads on FTU liquid lithium limiter

The choice of the best material exposed to the plasma in a future reactor is still an open question. One of main requirements to be satisfied is the capability to withstand high heat loads, in the range 10–20 MW/m², during normal operations in a future reactor, as well as the peak power released by ELMs in H-mode operation. On FTU, since the end of 2005, we have started an innovative program having as main goal the possibility to expose a liquid surface to the plasma. The small wetted area, of the FTU three liquid lithium limiter units, does not allow to use it as main limiter for all the duration of the discharge so that it is always set in the shadow of the main toroidal limiter. In this condition, heat loads up to 2 MW/m² are normally withstood by the liquid lithium limiter without any surface damage and problems to the FTU operations. In order to increase the heat load on the liquid lithium limiter for a controlled limited period, the plasma column is shifted towards the liquid lithium limiter during the discharge. The surface temperature remains constant although the plasma column is pushed on the liquid lithium limiter. This saturation of the surface temperature can be understood considering the dependence of the evaporation rate versus the surface temperature between 250 °C and 550 °C that increases by five orders of magnitude. The evaporated lithium forms a strongly radiative cloud all around the three units limiting the power load on the surface. We do not observe any accumulation of lithium into the discharge as it can be also inferred from the time evolution of the Li III line growing up until the temperature is reaching the maximum value and then remaining almost constant.     

Effect of transient heating loads on beryllium

Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m². Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process.     

Wetting properties of liquid lithium on select fusion relevant surfaces

Research into lithium as a plasma facing component material has illustrated its ability to engender low recycling operation at the plasma edge leading to higher energy confinement times. Introducing lithium into a practical fusion device would almost certainly require the lithium to be flowing to maintain a clean lithium surface for gettering. Several conceptual designs have been proposed, like the LiMIT concept of UIUC (Ruzic, 2011). Critical to the implementation of these devices is understanding the interactions of liquid lithium with various surfaces. For a device that relies on thermoelectric magnetohydrodynamic drive, such as the LiMIT concept, two of the critical interactions are the wetting of materials by lithium, which may be characterized by the contact angle between the lithium and the surface, and the relative thermopower between lithium and potential substrate materials.Experiments have been performed into the contact angle of liquid lithium droplets with various surfaces, as well as methods to decrease the contact angle of lithium with a given surface. The contact angle, as well as its dependence on temperature was measured. For example, at 200 °C, tungsten registers a contact angle of 130°, whereas above its wetting temperature of 350 °C, the contact angle is less than 80°. Glow discharge cleaning of the target surface as well as evaporation of a thin layer of liquid lithium onto the surface prior to performing wetting measurements were both found to decrease the wetting temperature.     

Recent contribution of JET to the ITER physics

In recent years the JET scientific programme has focussed on addressing physics issues essential for the consolidation of design choices and the efficient exploitation of ITER in parallel to qualifying ITER operating scenarios and developing advanced control tools. This paper reports on recent achievements in the following areas: mitigation of edge localised modes (ELMs), effects of toroidal field (TF) ripple, advanced tokamak scenarios, material migration and fuel retention. Active methods have been developed to mitigate ELMs without adversely affecting confinement. A systematic characterisation of the edge plasma, pedestal energy and ELMs, and their impact on plasma-facing components as well as their compatibility with material limits has been performed. The unique JET capability of varying the TF ripple from its normal low value δB_T = 0.08% up to δB_T = 1% has been used to elucidate the role of TF ripple on confinement and ELMs. Increased TF ripple in ELMy H-mode plasmas is found to have a detrimental effect on plasma stored energy and density, especially at low collisionality. The development of ITER advanced tokamak scenarios has been pursued. In particular, β_N values above the ‘no-wall limit’ (β_N ～ 3.0) have been sustained for a resistive time. Gas balance studies combined with shot-resolved measurements from deposition monitors and divertor spectroscopy have confirmed the strong role of fuel co-deposition with carbon in the retention mechanism through long-range migration and also provided further evidence for the important role of ELMs in the material migration process within the JET inner divertor leg.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

Simulation of residual thermostress in tungsten after repetitive ELM-like heat loads

Brittle destruction of tungsten armour under action of edge localised modes of plasma instabilities (ELMs) in ITER is an important issue determining the lifetime of the divertor. Besides, cracking of the armour produces tungsten dust with characteristic size of 1–10 μm flying from the armour surface with velocities up to 10 m/s. Influx of the tungsten dust into the ITER confinement decreases the temperature of the plasma, reduces the thermonuclear gain and even may run the confinement into disruption.This paper describes experiments in QSPA-Kh50 plasma gun and modeling, which has been performed for providing more insight into the physics of tungsten cracking under action of ELMs and for confirmation of the important result on stabilization of the crack development at the tungsten armour surface, predicted in our previous paper – the same authors, 2010.The threshold value of energy density deposition for start of tungsten cracking has been measured as 0.3 MJ/m² after 5–10 shots. From analytical considerations three times smaller threshold value has been predicted with increasing number of shots.     

Spatially-resolved flat-field soft X-ray spectrometer on experimental advanced superconducting tokamak

A space- and time-resolved flat-field soft X-ray spectrometer with the wavelength range of 1–13 nm has been developed to study impurity behavior on the Experimental Advanced Superconducting Tokamak (EAST). Using an entrance slit, a varied line spacing grating (2400 grooves/mm at the grating center), and a charged coupled device (CCD) system, time evolution of profiles of impurity line emissions were recorded. The spectral resolution of the spectrometer is 0.006 nm at 5 nm when the width of entrance slit is set at 0.03 mm. The best spatial resolution obtained is 24.5 mm with the height of slit at 1.0 mm. The spectrometer is placed 8000 mm away from the plasma center and the observed spatial range covers 0–450 mm from the equatorial plane of EAST. The first experimental results were obtained from the recent EAST campaign. The system was shown to be capable of observing spectral lines from both intrinsic low-Z impurities (C, O, et al.) and highly ionized medium- and high-Z impurities (Fe, Cr, Ni, Cu, et al.). Spectral lines from the full wavelength range (1–13 nm) can be obtained by moving the position of the CCD. Spectra with the wavelength intervals of 1–2 nm show strong metal lines for H-mode discharges. Time evolutions of C VI (3.373 nm) and O VIII (1.897 nm) lines are presented and detail analysis is performed combining electron density intensity, D_α and soft X-ray and extreme ultraviolet (XUV) radiation intensities. Evolutions of profiles of C VI (3.373 nm) and O VIII (1.897 nm) at core plasma were also shown, indicating that the spectrometer can be applied for impurity transport studies,     

Recent progress of NSTX lithium program and opportunities for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six 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 dual lithium evaporation system which can evaporate up to ～160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER.