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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.

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.

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.

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.

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.

It is well known that the radiation hardening or embrittlement of pressure vessel steels is very sensitive to the contents of minor solutes. To study the effect of dislocation loop formation on radiation hardening in these steels, in situ observation using a high-voltage electron microscope was conducted for the reference pressure vessel steel JRQ and Fe-based model alloys containing Mn, Si, and Ni. In the Fe-based model alloys, the addition of Mn was most effective for increasing dislocation loop density at 290 °C. Based on the assumption that a di-interstitial was adopted as the nucleus for the formation of an interstitial loop, a binding energy of 0.22 eV was obtained for the interaction of a Mn atom and an interstitial. The formation of Mn clusters detected by three-dimensional atom probe and interstitial-type loops at room temperature clearly showed that the oversized Mn atoms migrate through an interstitial mechanism. The temperature and flux dependence of loop density in pressure vessel steels was very weak up to 290 °C. This suggests that interstitial atoms are deeply trapped by the radiation-induced solute clusters in pressure vessel steels.

Tungsten (W) samples, undamaged and damaged by irradiation with 2.4 MeV Cu2+, were placed flush to the first wall of Large Helical Device (LHD) nearby the divertor target, and were exposed to 134 shots of hydrogen plasma discharges. Thereafter, to evaluate the enhancement of hydrogen isotope retention for W due to LHD hydrogen (H) plasma exposures, the W samples with and without LHD H plasma exposures were implanted with 1.0 keV D 2 + . For W samples without LHD H plasma exposures, the D retention in the damaged W was by a factor of ∼6 higher than that in the undamaged W. After LHD H plasma exposures, the carbon layers ∼4 nm in thickness were deposited on the W surfaces. Due to trapping of D atoms in the carbon layers, the D retention enhancement for undamaged and damaged W samples exposed to LHD H plasmas was significantly higher than that for the W samples without LHD H plasma exposures.

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.

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.

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.

In future fusion devices, helium irradiation to tungsten could lead to the formation of tungsten nanostructure, which has both pros and cons as a plasma facing component. Although the nanostructure may have considerable influences in fusion devices, there still exist open questions with regards to the formation mechanism and its characteristics. Here we show from a detailed TEM investigation the roles of helium nanobubbles on the formation of the nanostructured tungsten. The result also indicates that the growth of helium bubbles beneath the nanostructured layer could lead to the formation of tungsten dusts, which might have fatal influence to the tokamak operation, when the surface temperature is significantly high, say higher than 1800 K.

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.

Nano-geological diagnosis has been applied to the deposition layer formed on the Si specimens installed at the first wall surface during 2007FY Large Helical Device (LHD) experiment. Cross-sectional views of the deposition layers with nanometer-resolution were successfully observed by using focused ion beam fabrication technique and transmission electron microscopy observations. On the specimen located close to the divertor plates, deposition layer with ∼800 nm thick was observed, and it has very fine and stratified layer structures in nanometer-level. Thickness of each layer is so various from 5 to 100 nm, and majority of the composition elements such as Fe, Cr, Ni, O and C were continuously changed towards the depth direction from the top surface. Characteristics of each layer show the operational history in the LHD and the mechanism of dusts formation by flaking of the layer. This work provides the completely new knowledge about the deposition layer in fusion devices.

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.

In order to investigate the morphology change of tungsten due to helium plasma exposure, tungsten (W) materials fabricated in different processes, ITER grade W, ultra-fine grained W (W–0.8TiC), and powder metallurgy W after helium plasma exposure were observed by scanning electron microscope. In all the three cases, pin-holes are observed on the surface; moreover, the non-uniformity of the formation of helium bubbles and fiberform nanostructure was observed on ITER grade W. From the detailed analysis of the crystal orientation with the help of focused ion beam method (FIB) indicates that the initial growth of the nanobubbles may be slow on the (1 1 0) face of tungsten.

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.

It is well known that radiation hardening and embrittlement of A533B steels is very sensitive to minor solute content. To investigate the effect on microstructural evolution of these materials, A533B steels with different Cu levels were irradiated by 2.4 MeV Fe ions. The ion irradiation was performed at 290 °C and also at room temperature up to a dose of 2 dpa. After the irradiation, the microstructure of the specimens was observed by transmission electron microcopy. Small dislocation loops formed in the matrix and also in the vicinity of dislocation were identified as interstitial dislocation loops, using additional electron irradiation at room temperature in the HVEM.

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.

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.

To investigate the effect of varying temperature irradiation on oxide dispersion strengthened copper alloy (GlidCop CuAl15), heavy ion irradiation was carried out by using 2.4 MeV Cu ions up to 30 dpa in the temperature range of 473–673 K. TEM observation of CuAl15 before irradiation showed high density of small alumina particles and dislocations. After 30 dpa at 673 K constant temperature irradiation, voids were formed. However, voids formation was strongly depended on grain size. On the other hand, by a varying temperature irradiation, the number density of voids was decreased as compared with that of 673 K constant temperature irradiation. The result of CuAl15 is consistent with our previous studies of pure copper irradiated under same irradiation conditions.

Two types of plasma spray tungsten coatings on ferritic/martensitic steel F82H made by vacuum plasma spray technique (VPS) and air plasma spray technique (APS) were examined in this study to evaluate the possibility as plasma-facing armor. The VPS-W/F82H showed superior properties. The porosity of the VPS-W coatings was about 0.6% and most of the pores were smaller than 1–2 μm and joining of W/F82H and W/W was fairly good. Thermal load tests indicated high potential of this coating as plasma-facing armor under thermal loading. In case of APS-W/F82H, however, porosity was 6% and thermal load properties were much worse than VPS-W/F82H. It is likely that surface oxidation during plasma spray process reduced joining properties. Remarkably, both coatings created soft ferrite interlayer after proper heat treatments probably due to high residual stress at the interfaces after the production. This indicates the potential function of the interlayer as stress relieve and possible high performance of such coating component under thermal loads.

Nano-sized Ar bubbles give negative influence on the fracture resistance and occurrence of superplasticity in ultra-fine grained (UFG) W–TiC compacts. In order to enhance deformability in UFG, Ar-contained W–TiC compacts, effects of TiC addition on the high-temperature deformation behavior were examined. W–TiC compacts with TiC additions of 0, 0.25, 0.5, 0.8 and 1.1 wt% were fabricated by mechanical alloying in a purified Ar atmosphere and hot isostatic pressing. Tensile tests were conducted at 1673–1973 K (0.45–0.54 Tm, Tm: melting point of W) at initial strain rates from 5 × 10−5 to 5 × 10−3 s−1. It is found that as TiC addition increases, the elongation to fracture significantly increases, e.g., from 3 to 7% for W–0 and 0.25TiC/Ar to above 160% for W–1.1TiC/Ar when tested at 1873 and 1973 K at 5 × 10−4 s−1. The flow stress takes a peak at 0.25%TiC and decreases to a nearly constant level at 0.5–1.1%TiC. The ranges of the strain rate sensitivity of flow stress, m, and the activation energy for deformation, Q, with TiC additions are 0.17–0.30 and 310–600 kJ/mol, respectively. The observed effects of the TiC additions on the tensile properties are discussed.

Deuterium retention behavior in highly oriented pyrolytic graphite (HOPG), poly-crystalline diamond, poly-crystalline SiC, sintered WC, and converted B4C were investigated to reveal tritium behavior in re-deposition and co-deposition layers. Such layers would contain carbon, when the first wall and/or divertor were made of graphite or carbon-containing materials. Furthermore, the employment of other materials such as tungsten, and first wall conditioning such as boronization would complicate the layers. No different deuterium trapping sites due to carbon from those in HOPG were found in all the samples, where two deuterium trapping processes were observed: hot atom chemical trapping of energetic deuterium by a dangling bond of carbon and thermochemical trapping of thermalized deuterium in a constituent atom vacancy surrounded by carbons. Additionally, the latter reaction could be easily counteracted by or competed with the other deuterium trapping reactions by constituent atoms.

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.

A cyclic helium beam with energy 18.7 keV has been used to irradiate stress relieved and recrystallized powder metallurgy tungsten up to a fluence of ∼10 23 He/m 2 . After the helium irradiations, the samples have been exposed to pulsed high heat loading by an electron beam with a heat flux of 1 GW/m 2 and duration of 1, 1.5 or 2 ms. Surfaces of irradiated samples become smooth as a result of the electron beam irradiation of 1 GW/m 2 for 1 ms, due to atom migration on the surface at high temperature. On the other hands, the surface of irradiated samples is partially melted and resolidified. However, smoothing does not occur around the melted area. Stress relieved tungsten is melted by irradiation of a heat flux of 1 GW/m 2 and duration of 1.5 ms, but recrystallized tungsten is not melted by the same heat loading.

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.

To study retention properties of plasma particles in tungsten exposed to divertor plasmas, material probe experiments were carried out by exposing tungsten specimens to long pulse helium or hydrogen divertor plasma in LHD. After the exposure, the quantitative analysis of the implanted helium and hydrogen atoms were conducted by using ion beam analytical techniques. Though the incidence energies were very low, remarkable retention of plasma particles were confirmed. In the case of helium divertor experiment, the specimen which located at the strike point (highest retention), the total retained helium atoms were estimated to be 1.2 × 10 20 He/m 2 , and depth profile of helium atoms reached to about 25 nm. It is the much deeper range than that of the calculated results of 200 eV-He + irradiation by TRIM-code. In the case of hydrogen, the total retention of hydrogen at the strike point was estimated to be 3 × 10 19 H/m 2 (lowest retention).