Search

The effect of hydrogen on the structural change of vacancies in tungsten is analyzed by molecular dynamics in order to clarify the interaction between vacancies and hydrogen in tungsten. Simulations are performed at different temperatures (573, 773, 1073 K) and with different numbers of hydrogen in the vacancies (0, 18, 36, 54). Evaluating (1) isopotential surface for the increase of total potential energy, (2) the root mean square deviation (RMSD) of tungsten atoms, (3) the root mean square fluctuation (RMSF) of tungsten atoms and (4) density distribution in radial direction, we found that the presence of a large number of hydrogen in a vacancy at each temperature caused changes in the structure. This fact partially supports the experimental fact that in the coalescence of two tungsten vacancies, the retention of hydrogen activates them.

In the Large Helical Device (LHD), the first deuterium plasma experiment was conducted in 2017. To investigate tritium migration in the LHD vacuum vessel, long-term material probes were installed on the first wall before the deuterium plasma experiment. After the experiment, the microstructure and amount of tritium remaining in each probe were analyzed. The results showed that a relatively large amount of tritium remained in the probes on the first wall, forming a thick deposition layer, rather than in the probes located in the erosion-dominant area. In the deposition layers on the probes, the dominant element is carbon, which can be generated on the divertor tiles made of graphite. The result of orbit calculation of the energetic tritons in the case of the standard magnetic configuration in the LHD showed that approximately 40% of the tritons generated by deuterium–deuterium fusion reactions were promptly lost mainly to the divertor. Thermalized tritons also flew to the divertor along with the background plasma. The divertor tiles, on which the tritons impinged, were eroded by the divertor plasma, and carbon atoms and tritiated hydrocarbon molecules were generated and deposited on the first wall. This can be the dominant mechanism of tritium retention in the first wall. Among the material probes located in the erosion-dominant area, the amount of tritium remaining in the probe on which the energetic tritons impinged was relatively large. The results of the tritium balance analysis show that the first wall is not the dominant reservoir of tritium in the LHD.

Poly-crystalline tungsten (W) samples were simultaneously irradiated with Helium (He) and Deuterium (D) ions using the triple-ion implantation device. He effect on D retention and transportation was studied using different combination of ion energies and He/D flux ratios in the simultaneous implantation. The experimental results show that D trapping at dislocation loops is significantly reduced in the case of 3 keV He+–3 keV D2+at He/D flux ratios over 0.6. D trapping by stronger trapping sites such as vacancies and vacancy clusters showed less dependence on the flux ratio. On the contrary, the D retention increases at each He/D flux ratio in the case of 3 keV He+–1 keV D2+compared to only D2+ implantation even the He/D flux ratio reaches a value of 1.0. TEM observations confirmed that dense dislocation loops are formed rather than He bubbles, which is responsible for the enhanced D retention in W. Keywords: Simultaneous implantation, D retention, Helium, Flux ratio, Transportation, Thermal desorption spectroscopy

Growth of helium bubbles near the surface of metals leads to various morphology changes when sufficient helium ions are continuously implanted on the surface. In this study, low energy (< 100 eV) helium plasma irradiations to various noble metals were conducted, and consequent surface morphology changes were analyzed. It was found that fiberform nanostructures were formed by the helium plasma irradiation on rhodium and ruthenium thin films, which were deposited by magnetron sputtering, and on platinum and iridium wires. Growth of fiberform structures were not identified on gold, silver, and palladium samples exposed to helium plasmas even though the irradiations were conducted at various surface temperatures, though pinholes were observed on palladium surface. We discussed the relation between the morphology changes and material properties., ファイル公開：2020-04-25

The purpose of this study is to elucidate helium (He) retention behavior in tungsten (W) under simultaneous He and hydrogen (H) irradiation. Polycrystalline-W was irradiated by He+ and H2+ simultaneously with the energy of 1.0 keV and 3.0 keV. He+ fluences were (0.5, 1.0, 10) × 1021 He+ m−2 and H2+ fluence was 1.0 × 1022 H+ m−2,respectively. After irradiation, He desorption behavior was investigated by high temperature thermal desorption spectroscopy (HT-TDS) in the temperature range of R.T.-1773 K. Micro-structure changes of W after irradiation were observed by TEM. It was found that simultaneous irradiation with different H2+ energy significantly changed He retention behavior. 1.0 keV H2+ suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H2+ facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H2+.

Influence of mixed material layer formation on hydrogen isotope retention in W exposed to 2014 Large Helical Device (LHD) experiment campaign was evaluated by Thermal Desorption Spectroscopy (TDS), X-ray Photoelectron Spectroscopy (XPS), Scanning Electron Microscope (SEM) and Transmission Electron Microscope (TEM). It was found that a lot of hydrogen isotopes were trapped by the carbon-dominated mixed-material layer deposited on the plasma facing materials. Most of He was also trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K, respectively. However, the hydrogen retention behavior for erosion dominated area was clearly different from those for deposition dominated area and typical Plasma Wall Interaction (PWI) area, where He bubbles were introduced near the sample surface, leading to the introduction of various types of trapping sites in W.

All the hydrogen isotope (H, D, T) simultaneous TDS (Thermal desorption spectroscopy) measurement system (HI-TDS system) was newly designed to evaluate all hydrogen isotope desorption behavior in materials. The present HI-TDS system was operated under Ar purge gas and the H and D desorptions were observed by a quadruple mass spectrometer equipped with an enclosed ion source, although T desorption was evaluated by an ionization chamber or proportional counters. Most of the same TDS spectra for D and T were derived by optimizing the heating rate of 0.5 K s−1 with Ar flow rate of 13.3 sccm.Using this HI-TDS system, D and T desorption behaviors for implanted or DT gas exposed tungsten samples installed in LHD (Large Helical Device) at NIFS (National Institute for Fusion Science) was evaluated. It was found that major hydrogen desorption stages consisted of two temperature regions, namely 700 K and 900 K, which was consistent with the previous hydrogen plasma campaign and most of hydrogen would be trap...

The objective of this study is to investigate the effects of neutron (n) irradiation on hydrogen isotope transport in tungsten (W). W samples were irradiated with neutrons in a fission reactor and then exposed to high flux deuterium (D) plasma at 563 K in a linear plasma device. The fraction of D release increased with increasing exposure time. In addition, the D retention in n-irradiated W increased in proportion to the square root of the plasma exposure time. These observations were explained by increase in the penetration depth of D with filling up displacement damages acting as strong trapping sites.

Pure tungsten (W) samples were respectively installed on the two positions of typical plasma wall interaction area (PI) and erosion dominated position (ER) on the first wall in the Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan. After the experiment campaign in 2014, these samples were picked up and thermal desorption spectroscopy (TDS) were applied to evaluate their desorption behavior of He which was implanted by He discharge. It was found that a carbon-dominated mixed-material layer with impurities, such as Fe, Cr, Mo, O and N, etc., was deposited layer by layer during the plasma exposure on the PI sample. On the other hand, W-C mixed layer was formed on the ER sample. The results showed that a large amount of He was trapped in the samples on both PI and ER, and the total He retention for ER is about twice as large as that of PI. He was trapped in various types of trapping sites in the ER sample and their desorption peaks were located at temperatures of about 425 K, 755 K, 1130 K and 1630 K. For PI sample, most of He was trapped in the carbon-dominated mixed-material layer and the corresponding desorption temperature was limited to be about 600 K, 900 K and 1200 K. The additional 3.0 keV helium ion (He+) implantation was performed for several samples to investigate the He retention characteristics in these samples and it was found that no additional desorption stage was found. These results suggested that the He discharge history and its deposition on the W plasma facing wall would affect He desorption behavior of W.

This overview presents recent results regarding hydrogen isotope absorption and emission dynamics in neutron-irradiated tungsten (W) using our recently developed Compact Diverter Plasma Simulator (CDPS), a linear plasma device in a radiation-controlled area. Neutron irradiation to 0.016 - 0.06 displacement per atom resulted in a significant increase in deuterium (D) retention due to trapping effects of radiation-induced defects. We analyzed the dependency of D retention on the D plasma fluence by exposing neutron-irradiated pure W to D plasma at 563 K over a range of D fluence values. The total retention was revealed to be proportional to the square root of D fluence, indicating that the implanted D atoms first occupy the defects caused by neutron-irradiation near the surface and then the defects located in deeper regions. We further investigated the effects of post-plasma annealing on D emission; neutron-irradiated pure W was exposed to D plasma at 573 K and was then annealed at the same temperature for 30 hours. Approximately 10% of the absorbed D was released by annealing, suggesting that a heat treatment of the plasma-facing component of a fusion reactor at moderately elevated temperatures could contribute to the removal of accumulated hydrogen isotopes. The experimental results obtained in this study were only available by investigating neutron-irradiated specimens with the CDPS system, which will be essential for future studies of material behavior and plasma-wall interactions in the fusion reactor environment.

During the 19th experimental campaign in the Large Helical Device (LHD) in 2017, deuterium (D) plasma experiments were performed for the first time in the LHD. Tritons were generated by the D-D reaction in the core plasma during the plasma discharges. To remove the tritium in plasma facing surfaces for safety in maintenance, hydrogen (H) plasma experiments were performed for one month after the four months of D plasma experiments. In this study, the removal effect of the tritium by H plasma discharges was investigated. Six stainless steel (SUS316L) samples were introduced into a position of the first-wall of the LHD in the last month of the D experiments, which were exposed to the D plasma for one month and then, three samples were continued to exposure to the following H plasma for another one month. The experimental results showed that one-half of the tritium in the surface region of the SUS sample was removed, which was measured by the TIPT (Tritium imaging plate technique). Further, hydrothermal treatments were performed and the results showed that the tritium retention in the bulk of the SUS materials was also reduced. Meanwhile, the carbon content in the near-surface region was also reduced after H plasma experiments, which reduces the possibility of the formation of the C H chemical bond.

The Large Helical Device (LHD) started the deuterium plasma experiment on March 7, 2017. Approximately 6.4 GBq of tritium was produced in the first deuterium plasma experiment campaign and were utilized as tracers for the investigation of release behavior and the balance of tritium in the LHD vacuum vessel. To determine the tritium balance in LHD, the tritium release from the vacuum vessel was continually observed during the plasma experiment period and the vacuum vessel maintenance activities. The tritium exhaust rate was approximately 32.8% at the end of the plasma experiment. After the plasma experiment, the vacuum vessel was ventilated by room air for the maintenance activity and the tritium release from the in-vessel components was observed. The tritium release rate gradually decreased and became constant after four-month in spite of water vapor concentration. It is suggested that the tritium release mechanism from the vacuum vessel is a diffusion-limited process from the bulk. The tritium release amount during the maintenance activity for one year was approximately 5.0%. Considering the decrease of tritium decay for 1.5 years, tritium inventory in LHD was estimated to be approximately 3.66 GBq (57.2%) at the end of maintenance activity.

Hydrogen isotope exchange behavior on carbon (C)- tungsten (W) mixed material layer on W was studied as a function of gas exposure time and temperature. It was found that hydrogen isotope exchange enhanced the reduction of residual hydrogen effectively compared by only heating. The temperature dependence showed that hydrogen isotope exchange was the most effective at 523 K and above 573 K, thermal annealing reduced the total hydrogen isotope retention. It can be said that hydrogen isotope replacement is efficient to remove retained tritium on the surface region of mixture layer.

In the stellarator Wendelstein 7-X with its twisted 3D magnetic field geometry, studies of material migration with respect to first wall components become very important in view of the envisioned long-pulse operation. A variety of erosion/deposition samples were installed on the plasma-facing components exposed at three different nominal heat load levels between 0.1 and 10 MW/m2. After the first successful operation phase in divertor configuration, all the probes at higher and lower load levels were removed, whereas at the intermediate load levels, 352 out of 30 000 screws have been exchanged at selected locations along the toroidal and poloidal directions. The exchanged probes have been analyzed by various measurement techniques. At the higher load levels where the probes were installed within the divertor, heavy erosion has been observed presumably at the strike line positions. Both, erosion and deposition phenomena have been found on the screw heads. The optical reflection measurement profile of the whole plasma vessel show the deposition patterns at similar locations in all the five modules. At the low load level, the Si-wafer probes are under investigation.

A tungsten (W) deposition layer is formed while a sample is exposed to a helium (He) plasma at the surface temperature of 473 or 573 K. The formed He-W co-deposition layer was composed of fine grains, the size of which increased with the surface temperature. The samples with the co-deposition layer was exposed to a deuterium (D2) plasma, and the D retention was investigated using thermal desorption spectroscopy. It was found that the co-deposition layer has quite different D retention characteristics compared with bulk W.

The generation of tungsten fuzzy nanostructure by exposure to helium plasma is one of the important problems for the use of tungsten material as divertor plates in nuclear fusion reactors. In the present paper, the formation mechanisms of the helium bubble and the tungsten fuzzy nanostructure were investigated by using several simulation methods. We proposed the four-step process which is composed of penetration step, diffusion and agglomeration step, helium bubble growth step, and fuzzy nanostructure formation step. As the fourth step, the formation of the tungsten fuzzy nanostructure was successfully reproduced by newly developed hybrid simulation combining between molecular dynamics and Monte-Carlo method. The formation mechanism of tungsten fuzzy nanostructure observed by the hybrid simulation is that concavity and convexity of the surface are enhanced by the bursting of helium bubbles in the region around the concavity.

Processes occurring on the plasma-facing walls in the Large Helical Device (LHD) and in-vessel material migration were investigated using a technique of material deposition probes. Ten Si plates were located on the outer side of the first-wall surface in each 36° toroidal section (Nos. 1–10). Mixed-material layers deposited on the probes during the fiscal year 2012 plasma campaign contained carbon, boron, hydrogen, and metals from which the plasma-facing elements were made. Metallic impurities and hydrogen content in deposited layers were examined by Rutherford backscattering spectrometry (RBS) and elastic recoil detection analysis (ERDA), respectively. The cross-sectional observations of the deposited mixed-material layers were performed with the help of a scanning electron microscope equipped with an energy dispersive X-ray (EDX) spectrometer. The results of the combined RBS/ERDA analysis and TEM-EDX examination allow tracing the processes occurring on plasma-facing components during the plasma campaign.

We have developed a compact divertor plasma simulator (CDPS) that can produce steady-state deuterium and/or helium plasmas with densities above ∼ 1018 m−3 for Plasma-Wall Interaction (PWI) studies of neutron-irradiated materials. The maximum particle flux is about 1022 m−2s−1. The CDPS was installed and is being operated in the radiation-controlled area of the International Research Center for Nuclear Materials Science, Institute for Materials Research, Tohoku University. We are able to control sample temperature within uncertainty of 5 °C during plasma exposure by adjusting the cooling air flow rate to the sample holder. The CDPS has a sample-carrier system, which makes it possible to transfer a plasma-irradiated sample from the sample holder to an infrared heater for analysis using thermal desorption spectroscopy (TDS) without exposing it to the air. This avoids the oxidation of the sample and minimizes the time between the end of plasma exposure and TDS analysis. An ITER-like tungsten (W) sample (A.L.M.T. Corp.), which has been irradiated by neutrons to 0.06 dpa in a fission reactor, was exposed to a deuterium plasma in the CDPS. The experimental results clearly show that the total deuterium retention in the neutron-irradiated W sample increases significantly in comparison with a pristine W, as demonstrated by broadening of the TDS spectrum at high temperatures.

As the finalization of a hydrogen experiment towards the deuterium phase, the exploration of the best performance of hydrogen plasma was intensively performed in the large helical device. High ion and electron temperatures, Ti and Te, of more than 6 keV were simultaneously achieved by superimposing high-power electron cyclotron resonance heating onneutral beam injection (NBI) heated plasma. Although flattening of the ion temperature profile in the core region was observed during the discharges, one could avoid degradation by increasing the electron density. Another key parameter to present plasma performance is an averaged beta value $\left\langle \beta \right\rangle $ . The high $\left\langle \beta \right\rangle $ regime around 4% was extended to an order of magnitude lower than the earlier collisional regime. Impurity behaviour in hydrogen discharges with NBI heating was also classified with a wide range of edge plasma parameters. The existence of a no impurity accumulation regime, where the high performance plasma is maintained with high power heating >10 MW, was identified. Wide parameter scan experiments suggest that the toroidal rotation and the turbulence are the candidates for expelling impurities from the core region.

A fiberform nanostructured layer is formed on a tungsten (W) surface using helium (He) plasma irradiation. The behavior of the nanostructures and He bubbles were observed during annealing via in situ transmission electron microscopy. Notable changes in the nanostructures occurred at 1223 K. A collapse of the nanostructures was observed along with a collapse of the He bubbles during annealing.

Nanostructured tungsten formed by the exposure to helium plasma in a linear plasma device was installed in the large helical device (LHD). After the exposure in a series of experiments in the 2012 fiscal year campaign in LHD, the samples were analyzed by scanning electron microscope (SEM), transmission electron microscope (TEM), and energy dispersion x-ray spectroscopy (EDX). It was found that part of the nanostructures was totally covered with carbon based material probably from divertor, while some other parts were eroded by sputtering. On the erosion dominant region, it was revealed that the head part of nanostructures was sputtered and the surface became rounded, but the nanostructures still remained near the surface. Optical reflectance of the material was measured, and it was found that the morphology changes increased the optical reflectivity up to ∼10% from typically less than 1%. The possibility and limitation of the nanostructured tungsten as a light absorber (viewing dump) are discussed.

It has been found recently that low-energy helium (He) plasma irradiation to tungsten (W) leads to the growth of W nanostructures on the surface. The process to grow the nanostructure is identified as a self-growth process of He bubbles and has a potential to open up a new plasma processing method. Here, we show that the metallic nanostructure formation process by the exposure to He plasma can occur in various metals such as, titanium, nickel, iron, and so on. When the irradiation conditions alter, the metallic cone arrays including nanobubbles inside are formed on the surface. Different from W cases, other processes than growth of fiberform structure, i.e., physical sputtering and the growth of large He bubbles, can be dominant on other metals during irradiation; various surface morphology changes can occur. The nanostructured W, part of which was oxidized, has revealed a significant photocatalytic activity under visible light (wavelength >700 nm) in decolorization of methylene blue without any co-catalyst.