An overview of lithium experiments on HT-7 and EAST during 2012

In 2012, lithium coating with an upgraded system on EAST, the first application of lithium granules injection for ELMs pacing on EAST, and the first flowing lithium limiter experiments on HT-7 have successfully been carried out and several new results were obtained. On EAST, it was found that both the Mo first walls and the C divertors were well coated by lithium and the lithium film coverage was increased up to 85%, which greatly contributed to the new achievements of EAST, especially stationary H-mode plasma over 30 s and long pulse plasma over 400 s. And at the same time, ELMs suppression by active lithium conditioning and ELMs pacing using lithium granules injection were demonstrated and reported for the first time on EAST. On HT-7, flowing liquid lithium limiters using the TEMHD concept and using a thin flowing film concept were also initially tested and some references were obtained for the future development. Those experiments show that lithium should be an important material for fusion devices. It could be used for wall conditioning, ELMs mitigation and also provide a self-recovery plasma facing components in future fusion devices.     

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.     

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.     

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.     

The ARIES-AT advanced tokamak,Advanced technology fusion power plant

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β_N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ_x = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β_N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.     

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.     

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.     

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,     

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.     

R&D and maintenance on graphite tile of divertor region at EAST

EAST, with full superconducting magnetic coils, had been designed and constructed to address the scientific and engineering issues under steady state operation. The in-vessel components are full graphite tiles as first wall had been operated successfully. In the experiment campaign of 2010, the H mode operation and 1 MA operation have been gotten on EAST. However, in some case, some of the graphite tiles of divertor region are damaged with the plasma parameter enhanced. As most of the damaged graphite tiles are in the divertor region, they are probably damaged by the electro-magnetic force of the halo current when the VDEs occur. The force of the halo current is re-estimated. The structure analysis has been done by the ANSYS software. From the analysis result. It can be obtained that the stress is larger than the allowable stress when the halo current on the graphite tile is larger than 2.7 kA. The tensile testing of the graphite also has been done. As the result, the graphite tiles are damaged when the forces are up to 2400 N. To deal with the problem, two proposes are accepted. In the one hand, the new type graphite material is used, whose tensile strength is up to 45 MPa. In the other hand, the structure of the graphite tiles is optimized.     

Aerosol formation and hydrogen co-deposition by colliding ablation plasma plumes of lithium and lead

In a high-repetition inertial fusion reactor, along with pellet implosions, the interior of target chamber is to be exposed to high-energy, short pulses of X-ray, unburned DT and He ash particles and pellet debris. As a result, wall materials will be subjected to ablation, ejecting particles in the plasma state to collide with each other in the center of volume. The interaction dynamics of ablation plasmas of lithium and lead, candidate first wall materials, has been investigated in the deposited energy density range from 3 to 10 J/cm²/pulse at a repetition rate of 10 Hz, each 6 ns long. The plasma density and electron temperature of colliding ablation plumes have been found to vary from the order of 10⁸–10¹³ 1/cm³ and from 0.7 to 1.5 eV, respectively. The formation of aerosol in the form of droplet has been observed with diameters between 100 nm and 10 μm. Also, hydrogen co-deposition has been found to occur particularly for colliding plumes of lithium, resulting in the H/Li atomic ratio from 0.15 to 0.27 in the hydrogen partial pressure range from 10 to 50 Pa.     

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.     

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.     

Effect of implanted Si concentration on the Si nanocrystal size and emitted PL spectrum

Implantation of Si⁺ in excess into SiO₂ followed by annealing produces Si nanocrystals (Si-nc) embedded in the SiO₂ layer, which can emit a strong photoluminescence (PL) signal. Several samples have been characterized by means of ellipsometry and transmission electron microscopy (TEM). For local Si concentrations in excess of ∼2.4 × 10²² Si⁺/cm³, the Si-nc diameter ranges from ∼2 to ∼22 nm in the whole sample, the Si-nc in the middle region of the implanted layer being bigger than those near the surface or the bottom of the layer. The depth distribution of the Si-nc agrees relatively well with the SRIM simulation as well as with the depth distribution of the n and k components of the complex refractive index. For SiO₂ layers thermally grown on a Si wafer, the PL spectrum is modulated by optical interference of the pump laser and of the light emitted by the Si-nc in this layer. The good agreement between the results of the model calculations and experimental measurements indicates that for low and moderate Si concentration in excess (<8 × 10²¹ cm⁻³) the PL light emitters are localized in a layer situated at the same depth as the Si-nc depth distribution. However, for a Si concentration in excess of ∼2.3 × 10²² cm⁻³, the depth distribution of light emitters is narrow and situated mostly in the first half (relative to the surface) of the Si-nc depth distribution. This observation indicates that the recombination of the electron–hole pair at the interfaces could be responsible for the emitted PL spectrum.     

First-principles study of hydrogen adsorption and permeation in reconstructed cubic erbium oxide surfaces

Understanding surface properties of Er₂O₃, especially in relation to adsorption and permeation of atomic hydrogen, is of considerable importance to the study of tritium permeation barriers. In this work, hydrogen diffusion pathways through the low-index (1 0 0), (1 1 0), and (1 1 1) surfaces of cubic Er₂O₃ have been calculated using density functional theory within the GGA (PBE) + U approach. The dependence of the effective U parameter on lattice constants, bulk moduli, and formation energies of Er₂O₃ has been investigated in detail. The energetics of hydrogen penetration from the surfaces to the solution site in bulk Er₂O₃ were defined using the optimum effective U value of 5.5 eV. For a low surface coverage of hydrogen (0.89 × 10¹⁴ H/cm²), a penetration energy of at least 1.7 eV was found for all the low-index erbium oxide surfaces considered. The results of the present study will provide useful guidance for future studies on modeling defects, such as grain boundaries and vacancies, in tritium permeation barriers.     

Kinetic Monte Carlo simulation of hydrogen diffusion on tungsten reconstructed (0 0 1) surface

A kinetic Monte Carlo (KMC) algorithm has been developed to study the hydrogen diffusion on tungsten reconstructed (0 0 1) surface in the temperature range 220–300 K. The hydrogen diffusion coefficients predicted by the developed KMC code match the experimental values very well at low hydrogen coverage of a fraction of monolayer. A diffusion coefficient formula as a function of temperature and hydrogen coverage was derived from KMC simulations. Due to the very low probability of hydrogen occupying the long bridge adsorption sites, the rates of hydrogen atom having 3 or 4 neighbors are found to be zero for hydrogen coverage much less than a monolayer, while the rates of hydrogen atom having 0–2 neighbors are linear with respect to the hydrogen coverage. The calculated average rates of hydrogen located at the LB sites are very close to zero for low hydrogen coverage. Hydrogen only starts to occupy the LB sites after almost all SB sites are occupied.     

Characteristics of surface oscillation on high speed liquid Li jet

In the International Fusion Materials Irradiation Facility (IFMIF), high speed liquid lithium (Li) wall jet will be used as target irradiated by two deuteron beams of 125 mA at 40 MeV. To obtain knowledge of Li flow behavior, we have been studying on the surface wave characteristics experimentally using the liquid metal Li circulation loop at Osaka University. In this present study, the characteristic of surface oscillation on high speed liquid Li jet were examined. The free surface oscillation of Li flow was measured by an electro-contact probe apparatus, which detects electric contacts between a probe tip and Li surface. It was installed at 175 mm and 15 mm downstream from the nozzle exit to see influence of the initial growth of surface waves. The wave height of free surface waves was obtained from contact signal. While at 15 mm region the flow surface is very smooth covered with small waves in amplitude, the surface waves are developed sufficiently at the 175 mm. In the case of the velocity of 15 m/s, the maximum wave height reaches 4.8 mm. Heat deposition was estimated on the target back-plate with using the present statistical wave data.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Thermal,structural, and radiological properties of irradiated graphite from the ASTRA research reactor – Implications for disposal

The release of Wigner energy from the graphite of the inner thermal column of the ASTRA research reactor has been studied by differential scanning calorimetry and simultaneous differential scanning calorimetry/synchrotron powder X-ray diffraction between 25 °C and 725 °C at a heating rate of 10 °C min⁻¹. The graphite, having been subject to a fast-neutron fluence from ∼10¹⁷ to ∼10²⁰ n cm⁻² over the life time of the reactor at temperatures not exceeding 100 °C, exhibits Wigner energies ranging from 25 to 572 J g⁻¹ and a Wigner energy accumulation rate of ∼7 × 10⁻¹⁷ J g⁻¹/n cm⁻². The shape of the rate-of-heat-release curves, e.g., maximum at ca. 200 °C and a fine structure at higher temperatures, varies with sample position within the inner thermal column, i.e., the distance from the reactor core. Crystal structure of samples closest to the reactor core (fast-neutron fluence >1.5−5.0 × 10¹⁹ n cm⁻²) is destroyed while that of samples farther from the reactor core (fast-neutron fluence <1.5−5.0 × 10¹⁹ n cm⁻²) is intact, with marked swelling along the c-axis. The dependence of the c lattice parameter on temperature between 25 °C and 200 °C as determined by Rietveld refinement for the non-amorphous samples leads to the expected microscopic thermal expansion coefficient along the c-axis of ∼ 26 × 10⁻⁶ °C⁻¹. However, at 200 °C, coinciding with the maximum in the rate-of-heat-release curves, the rate of thermal expansion abruptly decreases indicating a crystal lattice relaxation. The ¹⁴C activity in the inner thermal column graphite ranges from 6 to 467 kBq g⁻¹. The graphite of the inner thermal column of the ASTRA research reactor has been treated by heating to 400 °C for 24 h in a hot-cell facility prior to interim storage.