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Low-energy helium (He) plasma irradiations were conducted on ruthenium (Ru) and rhenium (Re), which have hexagonal close packed (HCP) crystal structures. Growth of linear shaped fiberform nanostructures were identified on the surfaces of the both metals after the He plasma irradiation. We also conducted He plasma irradiation while Re particles were deposited on tungsten substrate; 3-mm-thick large scale fiberform nanostructures were grown on the surface. The crystal orientation was analyzed using diffraction patterns of Re and Ru nanofibers together with detailed transmission electron microscope observations. It was found that the growth of linear nanofibers has a preferential crystal orientation in the growth direction and it is always in the c-direction of the HCP crystals. Potential growth processes and mechanisms are proposed based on the experimental observations., ファイル公開：2021-12-01

The W (tungsten) samples were placed at top, equator and bottom walls of QUEST (Q-shu University Experiment with Steady-State Spherical Tokamak) device and exposed 1238 shots of hydrogen plasma during 2016A/W (Autumn/Winter) campaign with normal wall temperature of 473 K (maximum temperature of 523 K). Thereafter, the surface morphology was evaluated by color measurement, TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). Thick deposition layers were formed on the samples placed at the equator and bottom walls. On the other hand, thin mixed material layer was deposited on the top wall, where large H (hydrogen) retention was observed, which would be caused by dynamic plasma wall interaction (erosion and deposition) with higher H flux. Low H retention was confirmed for bottom wall, where higher wall temperature without He discharge would contribute. The additional 1 keV D2+ was implanted into these samples and deuterium retention enhancement was estimated. It was clearly found that the irradiation damages would induce more deuterium trapping than the formation of C–D bond.

0.8 MeV and 6 MeV iron (Fe) ions were implanted into tungsten (W) to produce the irradiation damages with the various damage distributions. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed to evaluate the D retention behavior on damage distribution in W. The experimental results showed that the total D retentions were decreased by increasing the damage concentration introduced near the surface region by 0.8 MeV Fe ion implantation. The retention of D trapped by vacancy clusters and voids, which would be the stable trapping sites with higher trapping energies, were reduced, suggesting that the recombination of D atom into D2 on the W surface was enhanced due to D accumulation near the surface region. It can be said that the hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface.

Helium (He) plasma irradiation to tungsten leads to formation of He bubbles and fiberform nanostructures called fuzz on the surface. Although the fuzz growth condition and growth rate have been well investigated, deposition of W together with He ions can change the He effect significantly. For example, the deposition leads to large-scale structures such as nano-tendril bundles and mm-thick large-scale fiberform nanostructures. In this review, current understanding of deposition effects on He irradiation are summarized, and the influence to fusion devices is discussed mainly from the view point of arcing.

Large-scale mm thick fiberform nanostructures are grown when materials are subjected to high flux helium ions with tungsten particles (ions and neutrals) precipitation. Observations of the large-scale fiberform nanostructures (LFNs) were conducted by scanning electron microscope (SEM) and transmission electron microscope (TEM) analyses. In particular, this study focuses on membrane nanostructures. Membranes that contain He bubbles were always observed when LFNs covered the surface. We discuss the growth process of the membrane structures based on detailed observations. Keywords: Nanostructures, Helium, Tungsten, Transmission electron microscope

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

Fiberform nanostructured niobium (Nb) was fabricated by one step helium (He) plasma irradiation. He ion implantation formed He nano-bubbles on a Nb plate and led to formation of protrusions while migrating in Nb matrix; fiberform nanostructures (FN) were grown when the fluence became high (>1026 m-2 ). The necessary conditions for the formation of Nb FN were revealed to be the surface temperature range of 900-1100 K and the incident ion energy higher than 70 eV. The sample was oxidized at 573-773 K in an air atmosphere, and Pt nanoparticles were photo-deposited on the Nb2 O5 samples. The surface was analyzed by scanning electron microscope, transmission electron microscope, x-ray photoelectron spectroscopy, and ultraviolet-visible spectrophotometry. Photocatalytic activity of the fabricated materials was studied using methylene blue (MB) decolorization process. An enhanced photocatalytic performance was identified on FN Nb2 O5 substrate with Pt deposition.

To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W. Keywords: He in damaged W, Hydrogen isotope retention behavior, Damaged W, Fusion

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+.

To evaluate the damage depth profile effect on hydrogen isotope retention in tungsten (W), combination usage of 0.8 MeV and 6.0 MeV Fe ions were implanted into W with the damage concentrations between 0.03 and 0.1 dpa. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed with the flux of 1.0 × 1018 D+ m−2 s−1 up to the fluence of 1.0 × 1022 D+ m−2, and the D retention behavior was evaluated by thermal desorption spectroscopy (TDS). The experimental results indicated that 6.0 MeV Fe ion irradiation would introduce vacancies and voids into bulk that were clearly controlled by the damage concentration, and the voids would become the most stable D trapping sites. It was found that D de-trapping from irradiation defects at lower temperature would be enhanced by the accumulation of defect near the surface due to 0.8 MeV Fe ion irradiation.

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.

Deuterium (D) retention behavior for the sequential 6 MeV iron (Fe) and 10 keV carbon (C) implanted tungsten (W) were evaluated by thermal desorption spectroscopy (TDS) and β-ray-induced X-ray spectroscopy (BIXS) to understand the synergetic effect of defect formation and C existence on D retention behavior for W under various damage distribution profiles. The experimental results indicated that retention of D trapped by dislocation loops was controlled by 10 keV C + implantation. The D retention was reduced in the sequential Fe 2+ − C + implanted W with higher C + fluence in comparison to that with lower C + fluence due to the formation of C-W layer which suppressed D diffusion toward the bulk and dense defects at the surface which reduce effective D diffusion coefficient. On the other hand, the amount of D trapped by the defects in the deeper region than C + implantation region (50 nm) was increased due to the formation of dense defects by 6 MeV Fe 2+ implantation within the depth of 1.5 μm.

The annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D2 spectra for Fe damaged W with lower defect concentr...

In this study, we measured deuterium (D) gas-driven permeation through tungsten (W) foils that had been pre-damaged by helium ions (He + ). The goal of this work was to determine how ion-induced damage affects hydrogen isotope permeation. At 873 K, the D permeability for W irradiated by 3.0 keV He + was approximately one order of magnitude lower than that for un-damaged W. This difference diminished with increasing temperature. Even after heating to 1173 K, the permeability returned to less than half of the value measured for un-damaged W. We propose that this is due to nucleation of He bubbles near the surface which potentially serve as a barrier to diffusion deeper into the bulk. Exposure at higher temperatures shows that the D permeability and diffusion coefficients return to levels observed for undamaged material. It is possible that these effects are linked to annealing of defects introduced by ion damage, and whether the defects are stabilized by the presence of trapped He.

A new temperature controlled material probe was designed for the exposure of tungsten samples to helium plasma in the LHD. Samples were exposed to estimated fluences of ∼1023 m−2 and temperatures ranging from 65 to 600 °C. Transmission Electron Microscopy analysis allowed the study of the impact of He irradiation under high temperatures on tungsten micro structure for the first time in real-plasma exposure conditions. Both dislocation loops and bubbles appeared from low to medium temperatures and saw an impressive increase of size (factor 4 to 6) most probably by coalescence as the temperature reaches 600 °C, with 500 °C appearing as a threshold for bubble growth. Annealing of the samples up to 800 C highlighted the stability of the dislocation damages formed by helium irradiation at high surface temperature, as bubbles and dislocation loops seem to conserve their characteristics. Additional studies on cross-sections showed that bubbles were formed much deeper (70–100 nm) than the heavily damaged surface layer (10–20 nm), raising concern about the impact on the material mechanical properties conservation and potential additional trapping of hydrogen isotopes.

From the helium (He) plasma irradiations to tungsten performed in the Magnum-PSI device, the effects of deposition of metals on the helium-plasma induced fiberform nanostructures (fuzz) are discussed. It was found that fuzz was not formed at the center of the plasma cylinder if there were significant metallic impurities from the source. Deposition of metallic impurities (mainly molybdenum and copper) counteracted the growth of fuzz. In addition to the effects of metals from the source, we installed a sputtering source near the sample to replicate the deposition environment in fusion devices. The thickness of fuzzy layer was ∼ 7 µm, which was about five times greater than that without deposition, at the He flux of 1.3 × 10 26 m − 2 , suggesting that the growth rate of fuzz layer was significantly accelerated due to the deposition of tungsten.

The W (tungsten) samples were placed at Top, Equator and Bottom plasma facing walls of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to 754 shots of hydrogen plasma during 2017A/W (Autumn/Winter) campaign. Thereafter, their surface morphologies and chemical states were evaluated by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). The XPS results showed that a thick carbon layer about 3–18 nm has formed throughout the wall surface. Among them, the Bottom wall had the deposition layer with the thickness of 2 nm, which was thinner than the top wall, namely erosion-dominated. On the other hand, a thick C layer about 18 nm was deposited on the Equator wall. The additional 1 keV D2+ was implanted into these samples and the D (deuterium) retention enhancement was estimated. The D2 TDS (Thermal Desorption Spectroscopy) spectra for all the samples had two major desorption stages at 400 K and 650 K, namely the desorption of D trapped by irradiation damages and deposition layer. The erosion/deposition profile would be caused by wall position and plasma condition, like a current start-up experiment. The desorption temperature of H2 (hydrogen) was shifted toward higher temperature side compared to that exposed to previous plasma campaign (2016 A/W), suggesting that H was mainly accumulated in the deposition layer with forming C–H bonds.

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.

Morphology changes by helium plasma irradiation are investigated on four metals that have hexagonal close packing (HCP) crystal structures. Straight fiberform nanostructures were identified on rhenium. Detailed transmission electron microscopy (TEM) observations of the rhenium nanostructures revealed that crystal orientation alters on locations where fibers bent. On bending locations, well-grown nanobubbles were observed or the fibers became very narrow. Parameter dependences of morphology changes are revealed on three of them (cobalt, hafnium, and zirconium). It is shown that various morphology changes such as nanocones, pinholes, pillars, and fiberform nanostructures occur by the helium plasma irradiation. Based on the temperature desorption spectroscopy (TDS) analysis, we will discuss the behavior of He atoms and morphology changes.

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.

Establishment of effective tritium removal method was one of important issues for the development of fusion reactor from the view of fuel recycle and safe operation. The deuterium (D) removal efficiency in tungsten (W) by energetic hydrogen (H) ions under room temperature and baking under 623 K were studied by thermal desorption spectroscopy (TDS). Iron (Fe) damaged W with various damage level by 6 MeV Fe2+ was adopted to simulate neutron irradiation damages. To understand the D removal behavior, the desorption of D2 was measured in-situ by a quadrupole mass spectrometer (QMS) during H2+ implantation and baking. The in-situ results showed that the desorption of D2 started after H2+ implantation and became slowly with the increment of H2+ implantation time. After H2+ implantation, part of D trapped by dislocation loops, vacancy clusters and voids could be removed by hydrogen isotope exchange. However, the removal efficiency by hydrogen isotope exchange decrease obviously as the presence of irradiation damages. The D trapped by dislocation loops and vacancy clusters can be removed by baking with high efficiency. It is worth to note that the D trapped by voids cannot be removed by baking leading to the lower D removal efficiency for W with high damage level.

The effect of heating temperature on deuterium (D) retention behavior for helium (He+) / carbon (C+) implanted tungsten (W) was studied. It was found that D retention behavior for He+ implanted W was not limited by the size of the He bubbles. The microstructure observation showed that the large helium bubbles were formed near the surface for He+ implanted W at 1173 K, suggesting that the D retention was reduced by the growth of the helium bubbles. In addition, to evaluate the effect of implantation ion species at high temperature, D retention behavior for He+ implanted W at 1173 K was compared with that for C+ implanted W at 673 K. It is concluded that the D retention depends on ion species, which makes different kinds of damages like He bubbles for He+ implantation and vacancy-ion complex (voids) for C+ implantation.

W samples were irradiated at 300 K with 6 MeV Fe ion with damage concentrations in the range from 0.0003 to 1.0 displacements per atom (dpa) and then implanted at 300 K with 500 eV D ions to a fluence of 5 × 10 21 D/m 2 . Deuterium retention in the damaged samples was examined in situ by thermal desorption spectrometry (TDS). Simulation of the TDS spectra was performed using the Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code to reveal the binding energies for deuterium captured by the ion-induced defects. It has been shown that the deuterium TDS spectra consist of two or three peaks (depending on the damage concentration) at about 400, 600 and 800 K, and can be simulated by the HIDT simulation code with the use of hydrogen-trap binding energies of 0.65, 1.25, and 1.55 eV.

The deuterium retention behavior for the Vacuum Plasma Spraying (VPS) tungsten (W) coating was studied to demonstrate the tritium retention as a function of heating temperature. It was found that t...

The effect of carbon implantation for the dynamic recycling of deuterium, which demonstrates tritium recycling, including retention and sputtering, was investigated using in-situ sputtered particle...

Helium plasma irradiation on metal surfaces leads to the formation of metallic fuzzy nanostructures accompanied by the growth of helium bubbles in metals. The mechanism of the growth process, its impact for fusion devices, and potential application have been explored. Here we show enhanced growth of large-scale fuzz by precipitating additional metallic particles during helium plasma irradiation. The growth rate of the fuzzy structures became orders of magnitude greater than conventional fuzz growth; in an hour of irradiation, 1 mm-thick visible tungsten and molybdenum fuzzy fur structures covered a tungsten metal substrate. Additional precipitation of metallic ions breaks the bottleneck diffusion process; moreover, further acceleration in the growth rate could have occurred if the electric sheath shape was influenced by the grown structure and the electric field that formed around the structure started collecting ions.

To understand the fundamental processes of deuterium retention and desorption of irradiated plasma facing materials, thermal desorption of deuterium in pure W with and without ion irradiation up to 2 dpa by 2.4 MeV Cu2+ have been investigated. After the ion irradiation at room temperature, high density of small interstitial typed dislocation loops and also nano-voids due to cascade collisions were detected by TEM observation. Thermal desorption of spectra of D showed that fine vacancy clusters are formed even at low fluence and at room temperature irradiation, these defects act as the strong trapping center of deuterium in the temperature range of 660–900 K.

Helium plasma irradiation to metals could be a potential novel surface modification method in nanoscale. In this study, systematic helium plasma irradiations to titanium were conducted in low ion energy ( 1025 m−2) regime. From the surface analysis by scanning electron microscopy and transmission electron microscopy, various types of morphology changes, i.e. nanocone, pinhole, porous structure, and rough surfaces, were identified. The experimental conditions that control the various morphology changes are revealed from the aspect of the incident ion energy and the surface temperature. Based on the observations, it is thought that helium bubble growth, surface diffusion, and physical sputtering played important roles for the morphology change.

Deformation of fiberform nanostructure and the dynamic behavior of helium (He) bubbles in fuzz tungsten (W) during annealing have been investigated by means of in situ cross-section observation using transmission electron microscopy and He desorption rate observation using thermal desorption spectroscopy (TDS). Thermal recovery of the nanostructure, such as shrinkage and coalescence of fine structure, annihilation of He bubbles, and large desorption of He gas, occurred around 1073–1173 K. The activation energy of He was estimated from a TDS peak that appeared around 300–400 K by using the Kissinger–Akahira–Sunose model-free-kinetics method. In addition, the TDS results of fiberform nanostructured tungsten were compared with those of tungsten samples irradiated with a high-energy He ion beam.

Thermal desorption of deuterium (D2) from deuterium-ion (D2+)-irradiated Fe–9Cr was correlated with the microstructural evolution of the alloy during irradiation with 8-keV D2+ ions following annealing to determine the retention and desorption behavior of the implanted deuterium and to identify effective traps for them, particularly at high temperature. After irradiation at 573 K, a new desorption stage formed between 650 and 1100 K at higher fluences, and cavities were observed using transmission electron microscopy. The total amount of trapped deuterium following irradiation with a fluence of 3.0 × 1022 ions/m2 was 6.8 × 1017 D2/m2, or approximately 0.007%. These results indicate that the deuterium atoms recombined to form D2 molecules at the surfaces of the cavities.

Microstructures of vacuum plasma sprayed (VPS) and brazed W on NIFS-HEAT-2 were studied by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). During the VPS process, growth of grains perpendicular to the interface was prominent. The blocky Ti-oxides, which have a strong effect on the mechanical properties before and after irradiation, were dissolved in the interface region of NIFS-HEAT-2. Understanding the effect of heat treatment (namely, post-coating heat treatment) on W-coated materials is important to reduce the irradiation induced hardening effects on the materials. After a heat load of about 5 MW/m 2 (∼16 cycles), the surface of VPS-W has a very fine rough structure but no prominent cracks due to thermal heat loads were detected.

In this study, a small tensile test machine was inserted in the beam line of a tandem-type accelerator. After 2.4 MeV Fe 2+ ion irradiation, the microstructure and hardness changes of the specimens with and without applied stress were studied. Without stress, the formation of small interstitial-type dislocation loops was prominent in the matrix and also in the vicinity of dislocations in the Fe–1.4 wt.%Mn alloy. At room temperature, radiation-induced hardening was more prominent in samples with stress than in samples without stress. However, at 563 K, the effect of stress on hardness changes was minor. TEM observations showed that the applied stress reduced loop nucleation and enhanced loop growth to a degree corresponding to the microstructure at higher-temperature irradiation. This study revealed that the formation of interstitial-type dislocation loops enhanced by Mn addition was essential for irradiation hardening of these samples both with and without applied stress at higher dose levels.

The deuterium (D) retention for Vacuum Plasma Spray (VPS)–tungsten (W)/F82H was studied using two different implantation methods, namely D plasma exposure and D2+ implantation. The D retention for polished VPS–W/F82H after plasma exposure was found to be reduced compared to that for polycrystalline tungsten. Scanning electron microscopy (SEM) and transmission electron microscopy (TEM) observations indicated that porous structures around grain boundaries and the interface between VPS–W layers would be potential D diffusion paths, leading to low D retention. In the case of D2+ implantation, the shape of D2 TDS spectrum was almost the same as that for D plasma-exposed VPS–W/F82H; however, the D retention was quite high for unpolished VPS–W/F82H, indicating that most of D was trapped by the oxide layer, which was produced by the VPS process. The reduction of surface area due to the polishing process also reduces D retention for VPS–W/F82H. These results indicate that controlling the surface chemical states is important for the reduction of tritium retention for future fusion reactors.

Many metallic mirrors will be used for optical diagnostics in ITER. This study investigates the effects of low-energy helium (He) irradiation on the reduction in optical reflectance for molybdenum (Mo) and tungsten (W). When the surface temperature of the materials was sufficiently low, (i.e., lower than 700 K for W and 600 K for Mo), optical reflectance did not change; however, at higher temperatures, optical reflectance decreased significantly due to the He irradiation. Surface analysis indicated that the reduction in optical reflectance was caused by the morphology change, likely due to the formation of He bubbles. Based on the experimental results, the influences of He irradiation on the diagnostic mirrors in ITER are discussed.

The D 2 + fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660 K in the TDS spectrum for a sample implanted with D 2 + up to the fluence of 10 23 D + m −2 , which desorption stage was not observed the D 2 + implanted sample with the fluence less than 10 22 D + m −2 . The TEM observation showed that the highly dense voids were formed in tungsten by D 2 + implantation with the fluence of 10 23 D + m −2 , considering that the D would be trapped by voids. To understand the D trapping by voids in C + implanted tungsten, C + –D 2 + sequential implantation experiments at various C + implantation temperatures were performed. It was found that the amount of D desorbed around 560 K was increased by increasing the C + implantation temperature. The formation of the voids was observed with increasing the C + implantation temperature by TEM, indicating that the increase of D desorption around 560 K was caused by the formation of voids. However, the desorption temperature of D trapped by voids in C + implanted sample was lower than that in D 2 + implanted one. TEM observation and XPS measurement indicated that this difference was caused by the increase of void size and/or the presence of implanted carbon.

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.