Search

Abstract: 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 1073–1173K. The activation energy of He was estimated from a TDS peak that appeared around 300–400K 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. [Copyright &y& Elsevier]

Abstract: 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 10min 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. [Copyright &y& Elsevier]

Abstract: 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 1800K. [Copyright &y& Elsevier]

Abstract: 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 1273K. 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. [Copyright &y& Elsevier]

Abstract: 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 1020 Hem−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. [Copyright &y& Elsevier]

Tungsten (W) is one of the most promising candidates for plasma facing materials in the fusion reactor. Helium (He) in W can influence the retention of hydrogen isotopes. In the present study, W targets were simultaneously irradiated by He+-H 2 + or He+-D 2 + ion beams with the energies of 1 keV or 3 keV, at fixed temperatures in a range of room temperature (R.T.) to 1073 K. Mechanisms of He and hydrogen interaction in W were discussed, especially from the point of He retention, which was characterized by the high-temperature thermal desorption spectroscopy (TDS) and glow-discharge optical emission spectroscopy (GD-OES) measurements. It is found that He desorption shifts to a lower temperature range for the W simultaneously irradiated by 3 keV He+-1 keV H 2 + at 573 K, under He+ fluence up to 1 × 1022 He+m−2. Transmission electron microscope (TEM) observation and annealing treatment at the temperature of 873–1073 K show that the increased He uptake is caused by the formation of dislocation. Enhanced retention amounts for the hydrogen isotopes were also confirmed. Amounts of the dislocation loops introduced by the H 2 +-only irradiation can be reduced by annealing treatment at 873 K, while that introduced by He+ irradiation are quite stable, which grows larger at elevated temperatures. With an increase of H 2 + energy, Helium uptakes at both weak trapping sites and bubbles are increased, while the amounts of hydrogen retention are decreased. It suggests that hydrogen ion has a significant influence on the He trapping sites at the irradiation temperature up to 573 K, while the hydrogen retention is determined by the distribution of He bubbles and dislocation loops. • Helium was mainly trapped by dislocation loops in tungsten irradiated by 3 keV He+- 1 keV H 2 + at 573 K. • Helium concentration in tungsten under 3 keV He+- 1 keV H 2 + irradiation was decreased at 873 K. • Increased hydrogen isotope uptake in tungsten was found when helium was mainly trapped at dislocation loops. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

• In order to simulate damage distribution at operation condition, heavy ions were implanted into W. • HIDT simulation code was applied to evaluate the depth profile of D. • The total D retentions were decreased by the damage introduced near the surface region. • The hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface. 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 (D 2 +) 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 D 2 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. [ABSTRACT FROM AUTHOR]

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 D 2 + 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. [ABSTRACT FROM AUTHOR]

• The retention of deuterium (D) for 1 dpa Fe ion damaged W-10%Re with dynamic-annealing or post-annealing was investigated by TDS. • The irradiation defects in W-10%Re by room temperature irradiation made a large D retention. • At higher temperature irradiation, D retention for W-10%Re was reduced less than half for both dynamic-annealing and post-annealing cases. • For post-annealing, higher temperature is required to reduce D retention at 800–1000 K compared to that for dynamic-annealing. The deuterium (D) retention for 1 dpa Fe ion damaged W-10%Re with dynamic-annealing or post-annealing was investigated by TDS. Major D 2 desorption was observed at temperatures of 400, 650 and 850 K for 1 dpa damaged W-10%Re without annealing (room temperature). By dynamic-annealing above 573 K, the D 2 desorption stage at 850 K was clearly eliminated. On the other hand, it was found that post-annealing at a higher temperature of 1173 K was required to annihilate this desorption stage. This indicates that the D desorption from the irradiation defects introduced at room temperature (post-annealing) needs higher annealing temperature compared to the dynamic-annealing. The D retention for dynamic-annealed sample was reduced by the recovery of D trapping sites by annealing. [ABSTRACT FROM AUTHOR]

Abstract 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 H 2 + simultaneously with the energy of 1.0 keV and 3.0 keV. He + fluences were (0.5, 1.0, 10) × 10 21 He + m −2 and H 2 + fluence was 1.0 × 10 22  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 H 2 + energy significantly changed He retention behavior. 1.0 keV H 2 + suppressed the He bubble growth and no bubbles can be observed at room temperature. On the other hand, 3.0 keV H 2 + facilitated the formation of He bubbles and increased the He retention due to the additional damage introduction by energetic H 2 + . [ABSTRACT FROM AUTHOR]

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 (D 2 + ) implantation was performed with the flux of 1.0 × 10 18 D + m −2 s −1 up to the fluence of 1.0 × 10 22 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 ∼10 23 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

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 Cu 2+ 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. [ABSTRACT FROM AUTHOR]

Highlights: [•] Systematic helium plasma irradiations to titanium were conducted. [•] The ion energy was <100eV and the high fluence was >1025 m−2. [•] Various types of morphology changes occurred on titanium surface. [•] They can be categorized nanocone, pinhole, porous structure, and rough surfaces. [•] He bubble growth, surface diffusion, and physical sputtering are important. [Copyright &y& Elsevier]

Thermal desorption of deuterium (D2) from deuterium-ion ( )-irradiated Fe–9Cr was correlated with the microstructural evolution of the alloy during irradiation with 8-keV 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 573K, a new desorption stage formed between 650 and 1100K 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. [ABSTRACT FROM AUTHOR]

Abstract: Vacuum plasma sprayed tungsten (VPS-W) coating formed on F82H steel kept at about 873K in conventional plasma spray conditions has inhomogeneous texture; it is a mixture of disorganized areas composed of large unmelted grains, fine randomly oriented grains and pores, and well-ordered areas composed of columnar grains. Heat load tests indicate that elimination of the disorganized area is necessary to improve the heat load resistance of VPS-W. One can get W coating with a texture of homogeneous columnar crystal grains by eliminating the unmelted large particles. Optimization of the W powder size was also effective to reduce the number of randomly oriented fine grains and pores. [Copyright &y& Elsevier]

Abstract: The D2 + fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660K in the TDS spectrum for a sample implanted with D2 + up to the fluence of 1023 D+ m−2, which desorption stage was not observed the D2 + implanted sample with the fluence less than 1022 D+ m−2. The TEM observation showed that the highly dense voids were formed in tungsten by D2 + implantation with the fluence of 1023 D+ m−2, considering that the D would be trapped by voids. To understand the D trapping by voids in C+ implanted tungsten, C+–D2 + sequential implantation experiments at various C+ implantation temperatures were performed. It was found that the amount of D desorbed around 560K 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 560K 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 D2 + 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. [Copyright &y& Elsevier]

Highlights: [•] The carbon-dominant mixed-material layer was formed on the tungsten surface after 15th plasma exposure in LHD. [•] The largest enhancement of deuterium retention was found to be about 21. [•] The major deuterium desorption temperature was shifted toward higher temperature side. [Copyright &y& Elsevier]

Arcing is a long-standing plasma-surface interaction issue, and the issue is currently being revived. This paper assesses the impact of arcing in fusion devices based on the observations in JT-60U, the Large Helical Device (LHD) and the linear divertor simulator NAGDIS-II. To investigate the first initiation process of arcing, field emission currents from several tungsten samples are measured. It is shownthat the field emission current increases significantly after tungsten is exposed to helium plasmas. A postmortem analysis of JT-60U tiles reveals that arcing phenomena occurred on carbon baffle plates inside the vacuum vessel in JT-60U. From the observation of the arc trails recorded on the baffle plate, the amount of eroded materials is discussed. The arcing seems to occur frequently on the inner baffles rather than the outer baffles. From LHD, it is shown that the arcing can be initiated on nanostructured tungsten even without transient events. The erosion of tungsten by arcing will become an important issue in a fusion reactor, where helium fluence is significantly increased. From the experiments in NAGDIS-II, it is shown that arcing can be initiated even without transient heat load when the target voltage is low enough, e.g. -500V. Frequent initiation of arcing annihilates the nanostructure growth due to helium plasma irradiation on the surface. [ABSTRACT FROM AUTHOR]

In magnetized plasmas, the presence of a significant number of energetic electrons has been observed but quantitative characteristics of these electrons are proving difficult to investigate. A Langmuir probe offers a means to provide quantitative measurement of these energetic electrons that takes into account electron emissions (secondary electron emission and electron reflection) from the probe tips and sheath expansion around the probe tips caused by a considerable negative potential. In this paper, these effects are experimentally confirmed and an analytical means to measure energetic electron characteristics are proposed. An analysis of plasmas produced by a high frequency wave is then applied leading to the successful detection of an asymmetric flow of energetic electrons. The estimated electron temperature and current density were approximately 4-5 keV and 2-3 kA/m2. [ABSTRACT FROM AUTHOR]

Abstract: 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. [Copyright &y& Elsevier]

Abstract: The simultaneous implantation of 10keV C+ and 3.0keV 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 673K, 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. [Copyright &y& Elsevier]

Abstract: 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 (110) face of tungsten. [Copyright &y& Elsevier]

Abstract: 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 (Γ perm) of hydrogen through a 30μm thick Ni membrane heated at 412–575K has been measured in the spherical tokamak QUEST. Γ perm is being measured during the scans of different operating parameters like RF power (P RF), chamber pressure (P chamber), discharge widths (τ dis) and vertical magnetic field (B Z). 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 Γ perm i.e. permeation fluence (Q perm) and the time integrated H α intensity i.e. H α fluence (Q α). Q perm 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. [ABSTRACT FROM AUTHOR]

Abstract: To elucidate the C+ implantation effect on deuterium retention in tungsten, simultaneous C+ and 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. [Copyright &y& Elsevier]

To evaluate the optical reflectivity deterioration of first mirrors for plasma diagnostics, in situ measurements of the laser reflectivity have been performed under the irradiation with low-energy helium ions in stainless steel, Mo and W. Surface morphology and radiation damage in sub-surface region have also been examined using scanning electron microscope, atomic force microscope and transmission electron microscope. It is found that the deterioration of the reflectivity under the irradiation with helium ions is very severe and depends on the ion energy, fluence and wavelength. Although fine roughness on the specimen surface was observed in irradiated materials, the size is too tiny to explain the present degradation of the reflectivity by using a Bennett's equation. On the other hand, from the present observation a strong correlation between the deterioration of the reflectivity and the formation of radiation damages in the surface and sub-surface regions of the laser penetration depth is suggested. [ABSTRACT FROM AUTHOR]

The undamaged W (tungsten) and 0.3 dpa (displacement per atom) damaged W samples by 6 MeV Fe2+ (Fe ion) irradiation were installed in the first wall of QUEST (Q-shu University Experiment with Steady-State-Spherical Tokamak) device and exposed to H (hydrogen) plasma in 2018A/W (Autumn / Winter) or 2019S/S (Spring / Summer) campaign to evaluate the impact of damages and impurities on hydrogen isotope retention. The surface morphology and chemical states of constituent atoms were observed by TEM (Transmission Electron Microscope) and XPS (X-ray photoelectron spectroscopy). It was found that thick Al deposit were found for the samples in 2019S/S campaign, which would come from the insulating plate during the CHI (Coaxial Helicity Injection) discharge. The additional 1 keV D 2 + was implanted into both of these samples and D (deuterium) retention enhancement was evaluated by TDS (Thermal Desorption Spectroscopy). The downward ion toroidal drift changed the impurity deposition and damage profiles. In 2018A/W and 2019S/S, the D retentions for undamaged W samples had position independent, indicating that plasma would be well-controlled in this configuration. In case of Fe2+ damaged samples, irradiation damages clearly changed the D retention characteristics. [ABSTRACT FROM AUTHOR]

Abstract: 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. [Copyright &y& Elsevier]

Abstract: To develop novel PPARα ligands, we designed and synthesized several 3-{3-[2-(nonylpyridin-2-ylamino)ethoxy]phenyl}propanoic acid derivatives. Compound 10, the meta isomer of a PPARγ agonist 1, has been identified as a PPARα ligand. The introduction of methyl and ethyl groups at the C-2 position of the propanoic acid of 10 further improved the PPARα-binding potency. [Copyright &y& Elsevier]

R&D for a MW-class solid target composed of tungsten was undertaken to produce a pulsed intense neutron source for a future neutron scattering-facility. In order to solve the corrosion of tungsten, tungsten target blocks were clad with tantalum by means of HIP’ing, brazing and electrolytic coating in a molten salt bath. The applicability of the HIP’ing method was tested through fabricating target blocks for KENS (spallation neutron source at KEK). A further investigation to certify the optimum HIP conditions was made with the small punch test method. The results showed that the optimum temperature was 1500 °C at which the W/Ta interface gave the strongest fracture strength. In the case of the block with a hole for thermocouple, it was found that the fabrication preciseness of a straight hole and a tantalum sheath influenced the results. The development of a tungsten stainless-steel alloy was tried to produce a bare tungsten target, using techniques in powder metallurgy. Corrosion tests for various tungsten alloys were made while varying the water temperature and velocity. The mass loss of tungsten in very slow water at 180 °C was as low as 0.022 mg/y, but increased remarkably with water velocity. Simulation experiments for radiation damage to supplement the STIP-III experiments were made to investigate material hardening by hydrogen and helium, and microstructures irradiated by electrons. Both experiments showed consistent results on the order of the dislocation numbers and irradiation hardness among the different tungsten materials. Thermal-hydraulic designs were made for two types of solid target system of tungsten: slab and rod geometry as a function of the proton beam power. The neutronic performance of a solid target system was compared with that of mercury target based on Monte Carlo calculations by using the MCNP code. [Copyright &y& Elsevier]

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. [ABSTRACT FROM AUTHOR]

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

Co-deposition of wall material and ions from plasmas occurs on material surfaces in fusion devices. In this study, we focus on tungsten (W) deposition layer in helium (He) plasmas (He-W co-deposition) formed at surface temperatures of 473–773 K. Usually, He irradiation at the surface temperature below 1000 K does not lead to significant morphology changes; we will show from transmission electron microscope observations that the roughness of the co-deposition layer has a significant temperature dependence even when the temperature is less than 800 K. We conducted deuterium plasma irradiation to the He-W co-deposition layer. It was found that the deuterium retention on the He-W co-deposition layer increased significantly with increasing the formation temperature of the co-deposition layer. [ABSTRACT FROM AUTHOR]

Abstract: 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 700K for W and 600K 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. [Copyright &y& Elsevier]

Abstract: The deuterium (D) retention for Vacuum Plasma Spray (VPS)–tungsten (W)/F82H was studied using two different implantation methods, namely D plasma exposure and 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 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. [Copyright &y& Elsevier]

Abstract: 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 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 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. [Copyright &y& Elsevier]

Abstract: 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 473K to 773K and ion fluence was 1024 m−2. In addition, during ion irradiation atmospheric pressure was 0.5mTorr. 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. [Copyright &y& Elsevier]

Abstract: Tungsten (W) samples, undamaged and damaged by irradiation with 2.4MeV 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 1.0keV . 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 ∼4nm 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. [Copyright &y& Elsevier]

Abstract: 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 {001}, {112}, {101}, and {111} surface, but there were no clear wavy structure on the crystal grains of {103}, {102}, {407} and {203}. 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. [Copyright &y& Elsevier]

Abstract: To elucidate the simultaneous implantation effects on deuterium retention in polycrystalline tungsten, simultaneous 10keV C+, 3keV and 3keV 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. D2 TDS spectrum for C+– implanted tungsten was clearly different from the samples with only , –He+ and C+––He+ implantation, indicating that D was trapped by C as C–D bond. For –He+ implantation sample, He+ implantation would introduce irradiation damages and D retention increased. However, C+––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+––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. [Copyright &y& Elsevier]