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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.

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.

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

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.

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.

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...

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.

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.

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.

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.

D retention in a VPS W coated V alloy (V–4Cr–4Ti, NIFS Heat 2) was investigated by an ion beam irradiation with a HiFIT device. Irradiation temperature changed from 473 K to 773 K and ion fluence was 1024 m−2. In addition, during ion irradiation atmospheric pressure was 0.5 mTorr. It was found that ambient gas absorption was the dominant process for D retention. Ion irradiation could enhance gas absorption probably due to surface modification (oxide layer removal) and could produce ion induced trapping sites. Although these ion induced processes can be avoided by W coating layers, D retention is still significant for VPS-W coated V alloy. D trapping could be done mainly in V-alloy and its mechanisms could be related to some intrinsic defects in V alloy and formation of titanium hydride at elevated temperature.

The effect of crystallographic orientation on the morphology change of tungsten (W) has been investigated in the linear plasma device, NAGDIS-I by exposing high density helium plasma to ITER grade W. After helium plasma exposure, the surface of specimens was analyzed by scanning electron microscope and orientation imaging microscope. The internal structure was analyzed with transmission electron microscope after the sample was milled by focused ion beam procedure. A clear difference in the surface morphology was observed on different crystal grains. Regular wavy structure were observed on crystal grains of {0 0 1}, {1 1 2}, {1 0 1}, and {1 1 1} surface, but there were no clear wavy structure on the crystal grains of {1 0 3}, {1 0 2}, {4 0 7} and {2 0 3}. The experimental results indicate that the angle between a slip face and face of each grains are important to determine the helium induced surface morphology change of tungsten.

Dynamics of deuterium recycling, including retention and sputtering behaviors was studied for C+ implanted tungsten. The amount of deuterium trapped by irradiation damages was clearly increased in the C+ implantation sample because the irradiation damages in the C+ implanted sample were formed more than those in the only D 2 + implanted one. In addition, the deuterium diffusion toward the bulk would be refrained by the formation of W–C mixed layer, which would work as the deuterium diffusion barrier. The in situ sputtered particle measurement system has been established and revealed that the formation of hydrocarbons such as CD4 was directly observed during D 2 + implantation into the C+ implanted tungsten. In the lower deuterium fluence, the CD4 sputtering rate was enhanced with increasing the deuterium fluence. It was considered that the sputtering rate of CD4 would be controlled by the concentration of deuterium on the top surface of the W–C mixed layer.

Helium plasma irradiation and electron heating experiments were conducted using tungsten in the divertor simulator NAGDIS-II. Helium plasma irradiation to tungsten led to the formation of nanostructures on the surface, while the nanostructures were annihilated after the potential of the specimen was changed to positive for several 10 min so that electrons irradiated the sample without ion irradiation. The specimens were analyzed in detail by transmission electron microscope with the help of focused ion beam technique. It is revealed that the helium nano-bubbles still remained even after the nanostructures were disappeared from the surface. Porosity of the nanostructured tungsten was measured from the TEM images.

The simultaneous implantation of 10 keV C+ and 3.0 keV D 2 + into tungsten was carried out at elevated temperatures to establish the retention mechanism of energetic deuterium and carbon. Thermal desorption spectroscopy showed that the deuterium retention obviously decreased as the implantation temperature increased. All the desorption stages disappeared when the implantation was performed at 673 K, indicating no deuterium trapping as C–D bonds was processed in this temperature. The results of X-ray photoelectron spectroscopy and glow discharge-optical emission spectroscopy showed that a mixed carbon layer had been formed during the implantation, resulting from the carbon deposition and accumulation near the surface of tungsten during the implantation. The carbon layer would enhance the chemical sputtering with deuterium and reduce the deuterium retention, as the implantation temperature increases. During discussion, a simple retention mechanism has been proposed, which shows the importance of implantation temperature.

Hydrogen including a trace amount of tritium was loaded on the edge surface of an F82H rod. After the loading, the rod was held at 298 or 323 K to allow hydrogen diffuse in and release out. Tritium tracer techniques have been applied to determine hydrogen depth profiles and hydrogen release rates by using an tritium imaging plate technique and a liquid scintillation counting technique, respectively. The depth profiles were composed of a surface localized component within 200 μm of the surface and a diffused component extending over 1 mm in depth. The apparent hydrogen diffusion coefficients obtained from the depth profile of the diffused component are near the extrapolated value of the literature data determined at higher temperatures. The surface localized component, which is attributed to trapping at surface oxides and/or defects, was released very slowly to give apparent diffusion coefficients much smaller than those determined from the diffused component.

Behaviors of hydrogen isotope retention and damages in tungsten and SS-316 with simultaneous C+–D2+ implantation were compared to those with only D2+ implantation using X-ray photoelectron spectroscopy (XPS), Thermal desorption spectroscopy (TDS), glow discharge-optical emission spectroscopy (GD-OES) and Transmission electron microscopy (TEM). The total D retention for SS-316 with only D2+ implantation was about 45% as large as that for tungsten. The D retention for simultaneous C+–D2+ implanted tungsten and SS-316 clearly increased as a factor of 1.7, which is almost the same among these samples. The density of dislocation loops was enhanced by the simultaneous C+–D2+ implantation, indicating the D trapping site would be produced by C+ implantation. As for the D desorption temperature, small shift toward lower temperature side was found for SS-316 compared to tungsten, indicating the D trapping energy by dislocation loops and grain boundary for SS-316 is lower than that for tungsten.

To elucidate the simultaneous implantation effects on deuterium retention in polycrystalline tungsten, simultaneous 10 keV C + , 3 keV D 2 + and 3 keV He + implantation was performed at room temperature. He + /D + flux ratio dependence of deuterium retention indicated that He + implantation induced high deuterium retention even if the He + /D + flux ratio was low. D 2 TDS spectrum for C + – D 2 + implanted tungsten was clearly different from the samples with only D 2 + , D 2 + –He + and C + – D 2 + –He + implantation, indicating that D was trapped by C as C–D bond. For D 2 + –He + implantation sample, He + implantation would introduce irradiation damages and D retention increased. However, C + – D 2 + –He + implantation, the accumulation of C on tungsten was suppressed and the retention of D trapped by C clearly decreased. These facts indicated the D retention for C + – D 2 + –He + implanted tungsten was limited by He + implantation and most of D would be retained in the interstitial sites or irradiation defects under simultaneous implantation with He.

Particle retention and recycling in plasma fusion devices are generally associated with the diffusion of atomic hydrogen into the materials. The resulted permeation of atomic hydrogen is known as plasma driven permeation (PDP). This permeation may also be significant, even in the walls, which are not directly exposed to the plasma. Under similar conditions, the permeation flux (Γperm) of hydrogen through a 30 μm thick Ni membrane heated at 412–575 K has been measured in the spherical tokamak QUEST. Γperm is being measured during the scans of different operating parameters like RF power (PRF), chamber pressure (Pchamber), discharge widths (τdis) and vertical magnetic field (BZ). Simultaneously edge plasma density and spectral intensities of atomic (Balmer) lines and molecular (Fulcher) bands have been compared with the permeation measurements. A linear relationship has been established between the time integrated Γperm i.e. permeation fluence (Qperm) and the time integrated Hα intensity i.e. Hα fluence (Qα). Qperm also shows a strong relationship with the edge plasma density and various spectral fluences. The obtained results are discussed for exploring the applicability of the permeation probes in measuring the atomic flux near the first walls.

Tungsten (W) samples were exposed to helium (He) plasmas with precipitation of W particles. It was shown that visible large-scale fiberform nanostructures (LFNs) were formed on the surface. The height of the LFN increased exponentially with the irradiation time in the initial phase of the growth. The growth rate was totally different from that of the conventional W nanostructure (fuzz) growth by He plasma irradiation, where the thickness increased in proportion to the square root of time. The LFNs growth had a directionality in a millimeter global scale; the direction was always ∼45–60 degree rotated in clockwise direction from the magnetic field. The plasma flow measured spectroscopically near the sample was found to significantly increase when inserting a negatively biased sample, and the growth direction of the LFN was consistent with the direction of the plasma flow. Moreover, highly porous fiberform nanostructures or cauliflower-like structures comprised the deposition layer even when the He incident ion energy was lower than the threshold energy of fuzz growth (20–30 eV).

High heat flux experiments using a hydrogen–helium mixture beam have been carried out on powder metallurgy tungsten (PM-W) and ultra fine grain W–TiC alloy (W–0.5 wt%TiC-H 2 ). The energy of is 18 keV. Beam flux and heat flux at the beam center is 2.0 × 10 21 atoms/m 2 s and 7.0 MW/m 2 , respectively. Typical ratio of He/D ion is 0.25. Beam duration is 1.5–3 s and interval of beam shot start is 30 s. The samples are irradiated up to a fluence of 10 22 –10 24 He/m 2 by the repeated irradiation pulses. After the irradiation, surface modification by the irradiation and hydrogen retention, surface composition have been investigated. Surface modification by hydrogen–helium mixture beams is completely different from results of single beam irradiation. In particular, mixture beam irradiation causes remarkably high hydrogen retention.

We investigated the magnetic and electrical properties of a heavy-ion irradiated single crystalline iron film. A high quality Fe (001) film on MgO (001) was fabricated using the molecular beam epitaxy technique. The iron film was treated by 3.2 MeV Ni ion irradiation at room temperature using a tandem accelerator. Although the residual electrical resistivity increased by 0:9 � 10 � 8 � m after ion irradiation, the M-H curves did not change significantly. These results indicate that the formation of small irradiation defects such as sub-nanometer size vacancy clusters has little influence on magnetocrystalline anisotropy and the magnetization process of iron. [doi:10.2320/matertrans.MAW200907]

It is important to evaluate tritium behavior in tungsten deposition layers considering a long-term plasma operation. In this study, tungsten deposition layers were formed by deuterium or helium RF plasma sputtering. The release behavior of deuterium or helium from the layers were observed by a thermal desorption method. When a tungsten deposition layer does not contain oxygen, the retained deuterium is mainly released as D{sub 2}. When oxygen exists in the layer, the majority of deuterium is released as water vapor. Tungsten deposition layers have an amorphous structure and consist offline grain with size of 2-3 nm. Numerous bubbles are observed in the layers. A formation of tungsten deposition layer in a fusion reactor may make tritium control more difficult. (authors)

Polycrystalline molybdenum specimens kept at room temperature and 873 K has been irradiated by 1.2 keV- and 8 keV-He + up to 3 × 10 22 He + /m 2 . Optical reflectivity has been measured with a spectrophotometer for the wave length between 190 and 900 nm. Surface morphology and chemical analyses has been examined with a scanning electron microscope, an atomic force electron microscope and X-ray photoelectron spectrometer. Loss of metallic gloss and remarkable reduction of reflectivity (190–900 nm) occurs by the irradiation of only 1 × 10 22 He + /m 2 . Reflectivity decreases with increasing ion energy, irradiation temperature and dose of helium ions. At rather low dose, it is reasonable to suppose that the reflectivity reduction may be due to the surface roughening and fine bubble formation. With increasing dose, roughness became smaller, while reflectivity decreased further. The further reduction of the reflectivity at high dose is considered to be due to the porous structure with nano-scale helium bubbles.

Plasma-facing and high heat flux materials in a fusion reactor suffer two types of damage: displacement damage caused mainly by high- energy neutrons and surface damage, such as erosion, sputtering, and blistering, caused by hydrogen and helium plasma. Usually, these two kinds of damage are investigated separately. In the present study, multiplier effects of helium plasma and neutron irradiation on microstructural evolution in tungsten were investigated using computer simulations based on a rate theory. Neutron irradiation with 10 � 6 dpas � 1 at 873 K and a helium flux of 10 18 m � 2 � s � 1 with 10-keV energy were used as typical irradiation conditions. The effects of irradiation temperature and defect production rate on the interaction between helium and defects were also investigated. The simulation results revealed the rapid diffusion of helium in tungsten. The helium concentration was saturated in tungsten 0.067 mm thick within 0.01 s at 873 K, and the formation of helium- vacancy clusters was nearly uniform in the matrix except at the surfaces facing towards and away from the irradiation after 0.01 s. This meant that the 10-keV helium plasma could enhance formation of helium-vacancy clusters in the region without damage produced by helium. The concentration of 6He-V clusters reached a very high level (10 � 6 ) even after a 1-s irradiation with a defect production rate of 10 � 6 dpas � 1 at 873 K. The concentration of helium-vacancy clusters decreased with increasing irradiation temperature and decreasing defect production rate. The accumulation of helium and the formation of helium-vacancy clusters depended on not only the vacancy concentration, which was determined by the irradiation temperature and defect production rate, and helium concentrations, but also irradiation time.

In order to investigate the effect of stepwise change of irradiation temperature on pure copper, heavy ion irradiations under constant temperature and varying temperature conditions were performed. Specimens were irradiated up to 10dpa. In the temperature region of 473K ~ 673K, one-step temperature variation and periodic temperature variations were performed during irradiation. The resulting microstructures were compared with those induced by the irradiation at the constant temperature of 673K.

Low-energy and high-flux plasma exposures on tungsten surfaces result in blister/hole formation. Subsequent He plasma exposure at 2,000 K on the blister-rich surface forms a hole structure on the surface, including the blister surface, and some blisters collapse due to recrystallization and bubble formation on the blister skin. Grain ejections were observed after deuterium plasma exposure at 550 K on a W surface on which a hole structure had formed due to He pre-exposure at 1,600 K.

Helium glow discharge cleaning (He-GDC) is frequently applied to reduce impurities like oxygen on the plasma facing materials in many plasma confinement devices, including large helical device (LHD). However, there are few reports which evaluated the wall conditioning effect from the point of view of the microstructure of materials. In the present study, therefore, microscopic damage to metals exposed to He-GDC in LHD were examined. Large numbers of dislocation loops and very high density of bubbles with sizes of 2–20 nm were formed by exposing to He-GDC. In addition, impurity deposits composed of Fe and Cr, which seemed to be re-deposited after sputtering of the first wall (SUS316L), were formed on all specimens. These serious surface modifications seem to have a larger impact on plasma fuel recycling under main plasma discharges. Instead of restoring the wall, significant damage by He-GDC creates gas trapping sites.

Metallic specimens mounted on a collector probe head were inserted in the scrape-off layer and exposed to long pulse tokamak plasma in TRIAM-1M. Strong radiation damage (formation of interstitial loops) was caused by high energy (>2 keV) charge-exchange hydrogen neutrals. The surfaces of the probe specimens were covered by new material made of co-deposited in-vessel elements such as Fe, Cr, Mo, O and C. The deposited layer was often exfoliated by blistering. These results indicate that energetic hydrogen neutrals and the co-deposition of impurities considerably change the surface properties of plasma facing components and even introduce new impurity sources.

Surface modification and internal damage of tungsten irradiated with 8keV Helium (He) ions at 1273 K was observed by the complementary use of a scanning electron microscope (SEM), an atomic force microscope (AFM), and a transmission electron microscope (TEM). It was found that an irradiated surface was roughened by the formation of dense large bulges of 100-200 nm and small ones of a few nm in size. It was also observed that fine He bubbles ranging from a few to 100 nm were formed in the subsurface layer. These results indicate that roughening of a He + ion irradiated surface at a high temperature results from the formation of bubbles just beneath the surface. Their thermal migration and coalescence under irradiation play an important role in the formation of various sized bubbles.

Thermal desorption of deuterium (D) from 8-keV D-ion irradiated beryllium (Be) above room temperature was correlated with microstructural changes during irradiation and annealing to understand the underlying mechanism of retention and trapping of D. D bubbles are formed at all examined temperatures between 300K and 873K. Large roundish bubbles above 200nm are especially formed above 573K. They remain even after annealing up to 973K. Strong retention of D by the bubbles occurs for the irradiation up to 673K. 8 refs., 5 figs.

High-purity molybdenum single crystals with the antitwinning axial orientation were irradiated with 14 MeV neutrons at room temperature, 473 K and 673 K under low neutron fluences of (2.1-2.7)×10 21 n/m 2 using the Rotating Target Neutron Sources (RTNS-II), and then were deformed in tension at 290 K. The irradiations led to the inhomogeneous yielding accompanied by large yield drops. The active slip plane, which of the unirradiated specimen was the plane (101), was shifted more and more towards the maximum shear stress plane (211) with decreasing irradiation temperature. By the isochronal annealing, the pronounced yield drop disappeared above 973 K

Fe-16Cr-17Ni ternary alloy and those with addition of 0.024P, 0.1P and 0.1P + 0.25Ti were irradiated with fast neutrons in FFTF/MOTA at 793 K and 874 K to 33 dpa and 60 dpa. For the purpose of investigating helium production effect, two levels of {sup 10}B were doped for each alloy. The microstructures and their relation with tensile properties were examined. The yield stress change was correlated with the density of phosphide precipitates.