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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.

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

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.

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.

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.

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.

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.

International audience; Impact of the helium plasma exposure on the surface modification in tungsten and RAFM (Reduced Activation Ferritic/Martensitic) steel have been investigated on the linear plasma device PSI-2 assuming the condition of DEMO first wall. In tungsten, a nanoscale undulating surface structure, which has a periodic arrangement, is formed under low temperature conditions below fuzz nanostructure formation threshold ∼ 1000 K. Interval and direction of the undulation shows dependence on the crystal orientation. A large variation in surface level up to 200 nm has been observed among grains at a fluence of $_{3 ×10}$$^{26}$ He/m$^{-2}$ showing dependence of the surface erosion rate on the crystal orientation. The {100} plane in which the undulating surface structure is not formed shows the highest erosion rate. This significant erosion is due to the multistage sputtering through impurity. In RAFM steel, sponge-like nanostructure is developed and it grows with increasing helium fluence beyond 1 $\mu$m. In the sponge-like nanostructure, a composition change from the base material is observed in which the tungsten ratio increases while the iron ratio decreases showing differences in sputtering ratio depending on the atomic mass.

A CW phased-array antenna system for electron cyclotron/Bernstein wave heating and current drive (ECWH/CD, EBWH/CD) experiments was developed in the QUEST. The antenna was designed to excite an elliptically polarized pure O-mode wave in oblique injection for the O-X-B mode conversion scenario, and its good performance was confirmed at a high power level. Long pulse discharges with a plasma current of 10 kA and 15 kA were non-inductively attained for 37 s and 20 s, respectively, with only radio frequency (RF) power. Divertor configurationswere also obtained in the RF-sustained plasmas. A new operational window for sustained plasma current was observed in the high-density plasma with a higher RF incident power. Two new heating and current drive systems with an 8.56GHz klystron and a 28GHz gyrotron are being prepared to conduct CW EBWH/CD experiments in the high-density plasma. © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research.

To examine the resolution of isotope analysis of hydrogen with glow-discharge optical emission spectroscopy (GDOES), depth profiles of hydrogen and deuterium in a H-containing Ta/D-containing Ti/Ni layered structure were measured. The depth profiles of deuterium could be measured with sufficient resolution in the presence of relatively large amounts of hydrogen and vice versa . In addition, measurements of depth profiles of He implanted in W at room temperature were also performed with Ne plasma. The intensity of the He emissions was sufficiently high at a fluence of 10 20 He m −2 or higher. The depth profiles of He measured in this manner were in good agreement with the results of cross-sectional observations using a transmission electron microscope. Therefore, it was concluded that GDOES with Ne plasma is a promising technique for the depth profile analysis of plasma-facing materials and deposited layers formed on them.

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.

PWR type nuclear reactor pressure vessels are neutron irradiated with stress during reactor operation. Therefore, it is important to know the effect of irradiation embrittlement of the material with and without stress conditions. It is already known that radiation hardening of the materials controlled by dislocation density in many materials. In the pressure vessel steels and model alloys, dislocation loop density is very sensitive to solutes (namely, Mn, Ni, Si etc.) levels containing in the alloy. Formation of small dislocation loops is prominent without applied stress in the matrix and also in the vicinity of dislocation in Mn containing alloys. These dislocation loops were identified as interstitial type dislocation loops. In this study, Microstructure evolution of Fe-1.4 mass% Mn alloy was studied in high voltage electron microscope (HVEM) by using a tensile sample holder. With stress condition, I-loop density was suppressed. But growth of I-loop becomes prominent.

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.

Rhodium and ruthenium thin film coatings were conducted on tungsten samples by using a magnetron sputtering device; then, we irradiated He plasma to the samples in the linear plasma device NAGDIS-II (Nagoya Divertor Simulator). Fuzzy nanostructures were formed on rhodium and ruthenium samples when the surface temperatures were ∼950 and ∼1200 K, respectively. When the surface temperature was high, i.e., >1200 K, it was found that tungsten atoms diffused across the rhodium film and reached the film surface, and tungsten-fuzz was formed over the rhodium layer. From TEM analysis of ruthenium fibers, it was identified that there were thin parts in the fibers, and some fibers had no He bubbles. It was likely that fiber growth mechanism on ruthenium was different from the other metals.

Plasma irradiation was used to create nanostructured vanadium oxide with potential commercial and industrial applications. Morphology changes were induced at the nano- and micro-meter scale, accompanied by the growth of helium nanobubbles. Micrometer-sized pillars, cube-shaped nanostructures, and fuzzy fiberform nanostructures were grown on the surface; the necessary conditions in terms of the incident ion energy and the surface temperature for those morphology changes were revealed. Hydrogen production experiments using a photocatalytic reaction with aqueous methanol solution were conducted on the fabricated samples. Enhanced H2 production was confirmed with the plasma irradiated nanostructured sample that had been oxidized in air atmosphere. Photocatalytically inactive vanadium oxide exhibited a high photocatalytic activity after nanostructurization of the surface by helium plasma irradiation.

Plasma wall interaction in fusion reactors is very complicated phenomena, because plasma facing materials suffer heavy irradiation of plasma particles, neutrons, impurity deposition and high flux heat load synergistically. This article introduces fundamental process of hydrogen and helium irradiation effects which are very important to understand plasma wall interaction in materials and also introduces recent studies on nano-scale plasma-wall interaction in actual experimental plasma confinement devices.

To elucidate the C+ implantation effect on deuterium retention in tungsten, simultaneous C+ and D 2 + implantation was performed on tungsten. Deuterium desorption behavior for simultaneously implanted tungsten was compared with that the sequentially implanted tungsten. It was found that deuterium retention for the simultaneous case was lower than that for the sequential one, indicating that the sputtering of carbon and deuterium trapped by carbon would prevent the deuterium trapping. In addition, deuterium retention decreased with increasing C+/D+ ratio and deuterium trapping by carbon was not observed for high C+/D+ ratio. These facts indicate that the number of available traps in tungsten produced by simultaneous ion implantation would mainly control the deuterium retention in tungsten. However, for low C+/D+ ratios, deuterium trapping by carbon would be one of major trapping states.

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]

Effects of neutron irradiation on YAG laser welded V–4Cr–4Ti alloy were studied in JMTR (Japanese Materials Testing Reactor). The samples with and without post-weld heat treatment (PWHT) at 1073 K were irradiated in the temperature range of 673–873 K up to the dose of 0.45 dpa. After the irradiation, the microstructure and Vickers hardness of the welded samples were compared of the base metal, which were simultaneously irradiated at the same irradiation cycle.

Beryllides have remarkable characteristics besides their low density, such as high radiation resistance, high chemical stability, low hydrogen isotope retention and good high temperature mechanical strength especially for fusion applications. In order to have a practical usage for industrial applications, it is important to compile a set of quantitative data on such properties. It is also important to develop fabrication and processing paths to assure the processing of inherently brittle beryllides. The characterization and manufacturing technologies development have been carried out mainly on the Be12Ti interemetallic compounds at the Be-rich side of the Be–Ti binary system. In the present paper, up-dated results on several properties are described for the compound fabricated by hot isostatic pressing (HIP) and ingot metallurgy. Mechanical properties of the compounds having a duplex microstructure with neighboring phases are evaluated by compressive tests from room temperature to 1273 K. Radiation damage of the compound is preliminary studied by charged particle irradiation. Oxidation in air and the interaction with water vapor are evaluated. Thermal desorption of the deuterium is examined by using transmission electron microscopy (TEM) and thermal desorption spectrometry (TDS). Through these evaluations it seems that Be12Ti is superior as neutron multiplier with respect to pure Be metal.

High density tungsten is coated by vacuum plasma spraying technique (VPS) on tiles, 20 mm × 20 mm × 10 mm. Substrate materials are carbon/carbon composite (CX-2002U), isotropic fine-grained graphite (IG-430U). Thickness of the tungsten-coating layer is 0.5 mm. The tungsten-coated tiles are jointed by Ti brazing on the OFHC surface with a cooling tube. In addition, VPS-W coated TZM also has been brazed. Thermal response and thermal fatigue lifetime tests using an electron beam facility have been carried out on the tungsten coated mock-ups under the active cooling condition. Heat flux is changed from 1 to 10 MW/m2. The heat flux experiments have been carried out under the condition that the water flow velocity, pressure and temperature are 15.0 m/s, 0.5 MPa and 293 K, respectively. The use of high density VPS-W improves the performance of mock-ups under steady state heat flux condition. In addition, in the case of high density W coated CX-2002U on OFHC and W coated TZM on OFHC, it is demonstrated that the mock-up successfully withstood 100 cycles of heat loads at 10 MW/m2 at steady state.

Mechanical properties of low energy He irradiated 304SS have been investigated by means of tensile test and nano-identation hardness tests. Specimen size is 2.8 mm × 5.4 mm × 0.1 mm in gauge and 12.4 mm in total length. Specimens are irradiated with 8 keV helium ions at R.T., 573 K and 873 K up to the fluence of 3 × 10 21 He/m 2 . Tensile tests are performed on unirradiated and irradiated specimens at a strain rate of 3.33 × 10 −3 s −1 . In every irradiation condition, one specimen is tested to failure and the other two specimens are strained plastically to 10% and 20%. Formation of blisters with a diameter of about 500 nm is observed on the surface of tensile-tested specimens after R.T. irradiation. Some exfoliation of the blisters is also observed. When the irradiation temperature increases, a large number of cracks with a size of about several micrometers are formed on the surface together with exfoliation during the tensile test. The yield stress of specimens irradiated at R.T. and 573 K increases about 10% in comparison with that of the unirradiated specimen. The result of nano-indentation tests indicate that hardness near the surface increases depending on the irradiation temperature.

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.

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.