Deuterium ion-surface interactions of liquid-lithium thin films on micro-porous molybdenum substrates

Lithium has been utilized to enhance the plasma performance for a variety of fusion devices such as TFTR, CDX-U and NSTX. Lithium in both the solid and liquid states has been studied extensively for its role in hydrogen retention and reduction in sputtering yield. A liquid lithium diverter (LLD) was recently installed in the National Spherical Torus Experiment (NSTX) fusion reactor to investigate lithium applications for plasma-facing surfaces (PFS). Representative samples of LLD material were exposed to lithium depositing and simulated plasma conditions offline at Purdue University to study changes in surface chemical functionalities of Mo, O, Li and D. X-ray photoelectron spectroscopy (XPS) conducted on samples revealed two distinct peak functionalities of lithiated porous molybdenum exposed to deuterium irradiation. The two-peak chemical functionality noticed in porous molybdenum deviates from similar studies conducted on lithiated graphite; such deviation in data is correlated to the complex surface morphology of the porous surface and the correct “wetting” of lithium on the sample surface. The proper lithium “wetting” on the sample surface is essential for maximum deuterium retention and corresponding LLD pumping of deuterium.     

Modelling of hydrogen isotope retention in the tungsten divertor of EAST during ELMy H-mode

In this work, we study hydrogen isotopes (HI) inventory inside tungsten plasma-facing materials during high confinement mode discharges with repetitive edge localized modes (ELMy H-mode) based on the operating parameters of the EAST device, since tungsten is considered as the primary plasma-facing material and the ELMy H-mode is an important operation regime for EAST and future devices. The upgraded Hydrogen Isotope Inventory Processes Code (HIIPC) is applied with the incident depth profile provided by SRIM-2013 to make the study. The code is first verified by comparison with experimental measurements. The effects of the incident ion energy and ion flux on the retention are then studied. Finally, using the parameters obtained from EAST diagnostics, the HI retention inside the W divertor during ELMy H-mode is studied, which indicates the retained HI can be increased dramatically mainly due to ion-induced trap sites by ELMs.     

Sorption and desorption behavior of hydrogen isotopes from tungsten deposits caused by deuterium gas or deuterium plasma exposure

To understand the behavior of hydrogen isotopes in deposits formed on plasma-facing wall is an important issue for development of a fusion reactor. In this study, sorption/desorption behaviors of hydrogen isotopes when tungsten deposits were exposed to deuterium gas or deuterium plasma at 300 °C were investigated. Samples of tungsten deposits were produced by the sputtering method using hydrogen plasma. After deuterium gas exposure or deuterium plasma exposure, the desorption behavior of hydrogen isotopes from the deposit was observed by the thermal desorption spectroscopy method. It was found that not a small amount of deuterium is retained in tungsten deposit by not only the plasma exposure but also the gas exposure while the amount of hydrogen incorporated in the deposit during sputter-deposition process is reduced. The amount of deuterium retained in the deposit by the plasma exposure was larger than that by the gas exposure in the experimental conditions in this work. The amount of hydrogen left after deuterium plasma exposure was larger than that after deuterium gas exposure.     

Fuel retention study in fusion reactor walls by micro-NRA deuterium mapping

Nuclear Reaction Analysis (NRA) with a ³He ion beam is a powerful analytical technique for analysis of light elements in thin films. The main motivation for ³He focused beam applications is lateral mapping of deuterium using the nuclear reaction D(³He,p)⁴He in surfaces exposed to a tokamak plasma, where a lateral resolution in the μm-range provides unique information for fuel retention studies.At the microprobe at the Jo?ef Stefan Institute typical helium ion currents of 300 pA and beam dimensions of 4 × 4 μm² can be obtained. This work is focused on micro-NRA studies of plasma-facing materials using a set-up consisting of a silicon partially depleted charge particle detector for NRA spectroscopy applied in parallel with a permanently installed X-ray detector, an RBS detector and a beam chopper for ion dose monitoring. A method for absolute deuterium quantification is described. In addition, plasma-deposited amorphous deuterated carbon thin films (a-C:D) with known D content were used as a reference.The method was used to study deuterium fuel retention in carbon fibre composite materials exposed to a deuterium plasma in the Tore Supra and TEXTOR tokamaks. The high lateral resolution of micro-NRA allowed us to make a detailed study of the influence of topography on the fuel retention process. We demonstrated that the surface topography plays a dominant role in the retention of deuterium. The deep surfaces inside the castellation gaps showed approximately two orders of magnitude lower deuterium concentrations than in areas close to the exposed surface.     

Study of the tungsten sputtering source suppression by wall conditionings in the EAST tokamak

The steady fusion plasma operation is constrained by tungsten(W) material sputtering issue in the EAST tokamak. In this work, the suppression of W sputtering source has been studied by advanced wall conditionings. It is also concluded that the W sputtering yield becomes more with increasing carbon(C) content in the main deuterium(D) plasma. In EAST, the integrated use of discharge cleanings and lithium(Li) coating has positive effects on the suppression of W sputtering source. In the plasma recovery experiments, it is suggested that the W intensity is reduced by approximately 60% with the help of ～35 h Ion Cyclotron Radio Frequency Discharge Cleaning(ICRF-DC) and ～40 g Li coating after vacuum failure. The first wall covered by Li film could be relieved from the bombardment of energetic particles, and the impurity in the vessel would be removed through the particle induced desorption and isotope exchange during the discharge cleanings. In general, the sputtering yield of W would decrease from the source, on the bias of the improvement of wall condition and the mitigation of plasmawall interaction process. It lays important base of the achievement of high-parameter and longpulse plasma operation in EAST. The experiences also would be constructive for us to promote the understanding of relevant physics and basis towards the ITER-like condition.     

The first results of deuterium retention on EAST with a full graphite wall via particle balance

This paper introduces the first results of deuterium retention on the Experimental Advanced Superconducting Tokamak (EAST) using particle balance.In the fall 2010 EAST experiments with a full graphite wall,the average deuterium retention fraction was about 19％ (including disruptive shots) and 38％ (not including disruptive shots).Fuel retention for the short-and long-pulse discharge was different.The H-mode discharges had a slightly lower fuel retention than the L-mode discharges.However,it was observed that disruptions introduced outgassing from the wall.Wall conditioning,such as lithium coating,increases retention.     

Preliminary results of in situ laser-induced breakdown spectroscopy for the first wall diagnostics on EAST

Post-mortem methods cannot fulfill the requirement of monitoring the lifetime of the plasma facing components(PFC) and measuring the tritium inventory for the safety evaluation.Laserinduced breakdown spectroscopy(LIBS) is proposed as a promising method for the in situ study of fuel retention and impurity deposition in a tokamak.In this study,an in situ LIBS system was successfully established on EAST to investigate fuel retention and impurity deposition on the first wall without the need of removal tiles between plasma discharges.Spectral lines of D,H and impurities(Mo,Li,Si,...) in laser-induced plasma were observed and identified within the wavelength range of 500-700 nm.Qualitative measurements such as thickness of the deposition layers,element depth profile and fuel retention on the wall are obtained by means of in situ LIBS.The results demonstrated the potential applications of LIBS for in situ characterization of fuel retention and co-deposition on the first wall of EAST.     

Deuterium retention in tungsten exposed to low-energy, high-flux clean and carbon-seeded deuterium plasmas

Depth profiles of deuterium trapped in tungsten exposed to a low-energy (≈200 eV/D) and high deuterium ion flux (about 1 × 10²¹ D/m² s) in clean (We use the term ‘clean’ in quotation marks having in mind the impossibility to obtain absolutely clean plasma. In our case the conception ‘clean’ D plasma means the plasma without intentionally introduced carbon impurities.) and carbon-seeded D plasmas at an ion fluence of about 2 × 10²⁴ D/m² and various temperatures have been measured up to a depth of 7 μm using the D(³He, p)⁴He nuclear reaction at a ³He energy varied from 0.69 to 4.0 MeV. The deuterium retention in single-crystalline and polycrystalline W increases with the exposure temperature, reaching its maximum value at about 500 K (for ‘clean’ plasma) or about 600 K (for carbon-seeded plasma), and then decreases as the temperature grows further. It is assumed that tungsten carbide formed on the W surface under exposure to the carbon-seeded D plasmas serves as a barrier layer for diffusion and prevents the outward transport of deuterium, thus increasing the D retention in the bulk of tungsten.     

Characterization of dust collected in EAST after 2019 campaign

Dust presented in experimental advanced superconducting tokamak(EAST) with mixed plasmafacing materials has been collected and characterized for the first time. Dust at different positions in the vessel was collected by vacuum cleaner after the first experimental campaign in 2019. The shape, composition, and size of dust particles have been analyzed using different methods. About80% of the total number of dust particles have size between 20 and 80 μm, and most of dust particles are spherical, while schistose shape, columnar and irregular shape were also found.With the help of energy-dispersive x-ray spectroscopy different elements of dust have been identified, which is generally consistent with the different plasma-facing components in EAST.Both x-ray fluorescence and inductively coupled plasma emission spectrometer are complementary methods for measuring the dust composition quantitatively. It was found that the major components of dust were lithium dust in the form of lithium carbonate and lithium hydroxide, which is due to the routine lithium wall conditioning during EAST operation.     

NSTX plasma operation with a Liquid Lithium Divertor

NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the LLD, i.e., about two times that of no-lithium conditions. The role of lithium impurities in this result is discussed. Following the 2010 experimental campaign, inspection of the LLD found mechanical damage to the plate supports, and other hardware resulting from forces following plasma current disruptions. The LLD was removed, upgraded, and reinstalled. A row of molybdenum tiles was installed inboard of the LLD for 2011 experiments with both inner and outer strike points on lithiated molybdenum to allow investigation of lithium plasma facing issues encountered in the first testing of the LLD.     

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.     

Temporal and spatial dynamics of optical emission from laser ablation of the first wall materials of fusion device

Laser-induced breakdown spectroscopy(LIBS) has been developed to in situ diagnose the chemical compositions of the first wall in the EAST tokamak. However, the dynamics of optical emission of the key plasma-facing materials, such as tungsten, molybdenum and graphite have not been investigated in a laser produced plasma(LPP) under vacuum. In this work, the temporal and spatial dynamics of optical emission were investigated using the spectrometer with ICCD.Plasma was produced by an Nd:YAG laser(1064 nm) with pulse duration of 6 ns. The results showed that the typical lifetime of LPP is less than 1.4 μs, and the lifetime of ions is shorter than atoms at ~10~(-6)mbar. Temporal features of optical emission showed that the optimized delay times for collecting spectra are from 100 to 400 ns which depended on the corresponding species. For spatial distribution, the maximum LIBS spectral intensity in plasma plume is obtained in the region from 1.5 to 3.0 mm above the sample surface. Moreover, the plasma expansion velocity involving the different species in a multicomponent system was measured for obtaining the proper timing(gate delay time and gate width) of the maximum emission intensity and for understanding the plasma expansion mechanism. The order of expansion velocities for various species is V_C~+ V_H V_(Si)~+ V_(Li) V_(Mo) V_W.These results could be attributed to the plasma sheath acceleration and mass effect. In addition, an optimum signal-to-background ratio was investigated by varying both delay time and detecting position.     

Temporal and spatial evolution measurement of laser-induced breakdown spectroscopy on hydrogen retention in tantalum

Fuel retention measurement on plasma-facing components is an active field of study in magnetic confinement nuclear fusion devices.The laser-induced breakdown spectroscopy(LIBS)diagnostic method has been well demonstrated to detect the elemental distribution in PFCs.In this work,an upgraded co-axis LIBS system based on a linear fiber bundle collection system has been developed to measure the hydrogen(H) retention on a tantalum(Ta) sample under a vacuum condition.The spatial resolution measurement of the different positions of the LIBS plasma can be achieved simultaneously with varying delay times.The temporal and spatial evolution results of LIBS plasma emission show that the H plasma observably expands from the delay times of 0-200 ns.The diameter of Ta plasma is about 6 mm which is much less than the size of H plasma after 200 ns.The difference in the temporal and spatial evolution behaviors between H plasma and Ta plasma is due to the great difference in the atomic mass of H and Ta.The depth profile result shows that H retention mainly exists on the surface of the sample.The temporal and spatial evolution behaviors of the electron excited temperature are consistent with that of the Ta emission.The result will further improve the understanding of the evolution of the dynamics of LIBS plasma and optimize the current collection system of in situ LIBS in fusion devices.     

Enhancement of hydrogen isotope retention capacity for the impurity deposited tungsten by long-term plasma exposure in LHD

The stress relieved tungsten samples were placed at three positions, PI (sputtering erosion dominated area), DP (deposition dominated area) and HL (Higher heat load area) during 15th plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ～ 6700 shots of hydrogen plasma in a 15th long-term experiment campaign in LHD. Thereafter, the additional deuterium ion implantation to these tungsten samples was performed to evaluate the change of hydrogen isotope retention capacity in the samples by long-term plasma exposure. It was found that the carbon-dominant mixed-material layer with more than 100 nm thickness was formed on a wide area of the tungsten surface. The thicker mixed-material layer was formed on the DP sample, where the deuterium retention was about 21 times as high as that for pure W. The major desorption temperature of deuterium was shifted toward higher temperature side, which was comparable to the trapping characteristic of carbon or irradiation damages.     

Ablation of Fusion Materials Exposed to High Heat Flux in an Electrothermal Plasma Discharge as a Simulation for Hard Disruption

Electrothermal plasma sources operating in the confined controlled arc discharge regime produce heat fluxes in the range expected for hard disruptions in future large tokamaks. The radiative heat flux produced inside of the capillary discharge channel is from the formed high density (10²³–10²⁷/m³) plasma with heat fluxes of up to 125 GW/m² over a period of 100 μs, making such sources excellent simulators for ablation studies of plasma-facing materials in tokamaks during hard disruptions. Graphite, beryllium, lithium, stainless steel, tungsten, copper, and molybdenum are among the materials proposed for use in fusion reactors. Computational experiments with the ETFLOW code using heat fluxes between 10 and 125 GW/m² have shown low total erosion for the low-z materials Li, Be and C and higher erosion for high-z materials Fe, Cu, Mo and W. The time rate of material erosion for various ranges of heat fluxes shows increased erosion with time evolution over the 150 μs pulse length of the simulated disruption event. At the highest values of simulated heat flux, low-z materials were found to ablate almost identically. At all simulated values of heat flux, the ablation of high-z materials correlated positively with the z-number.     

Deuterium inventory in plasma facing materials by means of NRA: A microbeam probe approach

Particle retention in tokamak walls is a key issue for long time discharges in future thermonuclear fusion reactors. Plasma wall interactions drive the fuel retention through two major mechanisms: co-deposition with carbon produced by wall erosion and particle retention in wall materials. In this study, we report results obtained from the tokamak Tore Supra, from which two types of samples were analyzed by means of micro-NRA: (i) small pieces of deposited carbon layers were collected after cumulative discharges and deuterium contents were measured; (ii) carbon fiber composite (CFC) samples, immersed in the plasma during an experimental campaign were also analyzed. 3D deuterium elemental mapping demonstrated that deuterium can be trapped at depths much higher than usual implantation depths and deep local retention sites have been evidenced and localized.This study demonstrates that μNRA can be used for assessment of deuterium post-mortem inventory in tokamaks, both by measuring uniformly distributed deuterium in small fragments of deposited carbon layers and by locally describing deuterium 2D and 3D distributions in complex structures.     

Development of vacuum ultraviolet spectroscopy for measuring edge impurity emission in the EAST tokamak

The dominant wavelength range of edge impurity emissions moves from the visible range to the vacuum ultraviolet(VUV) range, as heating power increasing in the Experimental Advanced Superconducting Tokamak(EAST). The measurement provided by the existing visible spectroscopies in EAST is not sufficient for impurity transport studies for high-parameters plasmas. Therefore, in this study, a VUV spectroscopy is newly developed to measure edge impurity emissions in EAST. One Seya-Namioka VUV spectrometer(McPherson 234/302) is used in the system, equipped with a concave-corrected holographic grating with groove density of 600 grooves mm~(–1). Impurity line emissions can be observed in the wavelength range ofλ=50–700 nm, covering VUV, near ultraviolet and visible ranges. The observed vertical range is Z=-350–350 mm. The minimum sampling time can be set to 5 ms under full vertical binning(FVB) mode. VUV spectroscopy has been used to measure the edge impurity emission for the 2019 EAST experimental campaign. Impurity spectra are identified for several impurity species, i.e., lithium(Li), carbon(C), oxygen(O), and iron(Fe). Several candidates for tungsten(W) lines are also measured but their clear identification is very difficult due to a strong overlap with Fe lines. Time evolutions of impurity carbon emissions of CII at 134.5 nm and CIII at97.7 nm are analyzed to prove the system capability of time-resolved measurement. The measurements of the VUV spectroscopy are very helpful for edge impurity transport study in the high-parameters plasma in EAST.     

Primary Results of Lithium Coating for the Improvement of Plasma Performance in EAST

First lithium coating associated with ion cyclotron range of frequency (ICRF) plasma was performed successfully in EAST. Results in reduction of both residual impurity and deuterium in the vacuum vessel were obtained. Particularly the partial pressure of deuterium after the lithium coating was reduced by about a factor of 5. Impurity radiation in the plasma was reduced and electron temperature increased by about 50%. Moreover, reproducible plasma discharges with high parameters, such as higher plasma current and density, could be easily obtained. These results showed that plasma performance was improved. Even though only 2 g of lithium were injected, the effective lifetime of the Li film was raised up to 40 shots.     

Deuterium retention in SiC coated graphite by D2+ implantation

The doped graphite tiles bolted to the active cooling heat sink, made of GBST1308 (1% B₄C, 2.5% Si, 7.5% Ti) coated with SiC, are now being used as the only plasma facing material (PFM) for the EAST device since the campaign of 2008. From the plasma density and fueling point of view, it is important to study thoroughly the hydrogen isotope retention in this kind of SiC-coated doped graphite. D₂⁺ implantations into the SiC coated doped graphite were performed at Shizuoka University. The chemical states of Si and C were studied by means of X-ray photoelectron spectroscopy (XPS), and the thermal desorption behavior of deuterium was analyzed by thermal desorption spectroscopy (TDS). It was found that deuterium was trapped by both C and Si in the SiC coatings. In the previous studies, Oya et al. reported the deuterium retention behavior in polycrystalline β-SiC. In this paper, difference of retention behavior in β-SiC and SiC coating will be also discussed.     

Studies of Li sputtering and particle recycling in glow discharges and in TJ-II

Lithium sputtering is studied at TJ-II with lithiated, boron coated walls. As reported previously, the sputtering yield measured in H plasmas was found to be significantly lower than expected, based on laboratory experiments and TRIM code calculations. The edge temperature dependence of the yield could be consistent with either an energy threshold far exceeding the corresponding pure lithium threshold, or with a strong perturbation of the impinging ion energies. The material mixing effect, one of the candidates for explaining the observed behavior, was studied by depositing boron on the Li layer. Also, particle retention and release was investigated in H/He plasmas in this very low recycling scenario. The release of either species in the opposite plasma was studied in different plasma conditions as well as in glow discharge (GD) plasmas. In He GD plasmas, the I/V characteristics of lithiated stainless steel electrodes were found to be anomalous, in spite of the fact that the dependence of sputtering on the incident particle energy agreed reasonably with expectations. The possible implications of these phenomena for the interaction of reactor plasmas with lithium elements are addressed.