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Abstract: Depth profiles of deuterium trapped in graphite tiles not exposed to hydrogen discharges just before air ventilation of JT-60U were determined based on the D(3He,p)4He resonant nuclear reaction analysis. The samples were taken from the W-shaped divertor. The maximum depth of analysis is about 1.4×1024 carbon atoms/m2, corresponding to a linear depth of 16μm for the density of 1.7×103 kg/m3. The highest concentration (19at.%) was found at the surface of the inner divertor area, which was significantly higher than that exposed to H discharges at the same location. However in the deeper region, the deuterium concentrations without H discharges were lower than those with H discharges in 3 of 4 samples. These results indicate that hydrogen discharges can remove deuterium trapped only in the shallow region. [Copyright &y& Elsevier]

Abstract: H and D retention characteristics in the near top surface of the JT-60U plasma facing carbon tiles were investigated by means of elastic recoil detection analysis. From the result, H was dominantly retained within 1μm depth because of the H discharges for the wall cleaning carried out before the removal of sample tiles. (H+D)/C ratios were small: it was ∼0.07 at a maximum in the divertor samples except for the deposition on the dome outer wing tile which was ∼0.13, and it was ∼0.15 at a maximum in the first wall samples. The hydrogen concentration in the redeposited carbon layers is strongly influenced by the temperature. Consequently, (H+D)/C ratios at the observed low level are a result of the operation of JT-60U at wall temperatures of 300°C or higher. [Copyright &y& Elsevier]

Abstract: Deuterium depth profiles in graphite tiles of JT-60U have been determined by means of the D(3He,p)4He nuclear reaction. Tiles with net carbon deposition showed high deuterium retention. A D/C ratio of 0.07 was found in the outer dome wing tile. In the inner divertor tile covered with a thick deposition layer, the depth profile shows a broad peak in a deep region. This indicates that deuterium trapped in the shallow region can be replaced by hydrogen with hydrogen discharges due to isotope exchange, while deuterium trapped in the deeper region can not. Deuterium in the redeposited layers can be more easily removed than that in the substrate. The deuterium concentration in the outer divertor target tile was very low (D/C<0.003). The highest D/C ratio, 0.09, was found in the upper region of the first wall. [Copyright &y& Elsevier]

Abstract: Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/12C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium–carbon co-deposition. In the outer and inner divertor target tiles, the D/12C data were lower than 0.006. Additionally, the maximum (H+D)/12C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium–carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk. [Copyright &y& Elsevier]

It is important to measure retentions and depth profiles of deuterium and tritium on plasma facing materials to understand plasma-wall interactions. The nuclear reaction analysis (NRA) has been established using a deuteron beam with an adjustable energy. By means of the NRA, depth profiles of deuterium and tritium on the surface of JT-60 divertor graphite tiles irradiated with DD burning were obtained. The NRA method showed that the average concentrations of the deuterium retained in the surface of two different were 1.3 × 1016 and 0.3 × 1016 cm-2. α-particles from the T(d, α)n reaction were measured from a dome unit tile of JT-60U W-shaped divertor and the measured activity of tritium was about 18 kBq/cm2. The peak energy of the α spectrum obtained with the NRA was about 1.2 MeV. This result implies that energetic tritons were implanted in the divertor graphite tile. [Copyright &y& Elsevier]

Abstract: Neutron transport calculations have been carried out to assess the capability of zirconium borohydride (Zr(BH4)4) and zirconium hydride (ZrH2) as advanced shield materials, because excellent shields can be used to protect outer structural materials from serious activation. The neutron shielding capability of Zr(BH4)4 is lower than ZrH2, even though the hydrogen density of Zr(BH4)4 is slightly higher than that of ZrH2. High-Z atoms are effective in neutron shielding as well as hydrogen atoms. The combination of steel and Zr(BH4)4 can improve the neutron shielding capability. The combinations of (Zr(BH4)4 +F82H) and (ZrH2 +F82H) can reduce the thickness of the shield by 6.5% and 19% compared to (water+F82H), respectively. The neutron flux for Zr(BH4)4 is drastically reduced in the range of neutron energy below 100eV compared to other materials, due to the effect of boron, which can lead to a reduction of radwaste from fusion reactors. [Copyright &y& Elsevier]

Abstract: Dust generated by plasma–wall interaction is a potential source of tritium retention in a fusion reactor. Evaluation of the dust accumulation in the entire vacuum vessel is required to estimate the total amount of tritium retention, but it was particularly difficult to measure for plasma-unexposed areas behind the PFC structures, i.e. “shadow areas”. Dust samples were collected at 3, 5 and 2–4 different toroidal locations on the first wall, divertor surface and the exhaust route under the divertor in JT-60U, respectively. On the tile surface, large mass area density was found at the inner divertor and baffle, in particular, upper tiles compared to the lower target tile where the thick deposition layers were produced. Mass area density was significantly increased at the shadow areas, i.e. under the divertor structure such as the divertor and baffle tiles and the divertor dome. It was found that the poloidal distribution is relatively symmetrical in the toroidal direction within a factor of three. In comparison with the previous collection just before major change of the plasma operations, dust accumulation was increased both at the exposed and shadow areas due to change in the operating conditions. [Copyright &y& Elsevier]

Abstract: Erosion/deposition distribution and hydrogen isotope behavior in the JT-60U plasma-facing wall were investigated. Distribution of the tritium, which was produced by D–D nuclear reaction, was not correlated with erosion/deposition distribution. The tritium distribution can be explained by the distribution of high-energy tritium ion-implantation due to ripple loss. Deuterium distribution in the divertor region was different from the tritium distribution and not well correlated with the deposition. The highest D/C was ∼0.05 at the bottom of the outer dome wing, which is much less than that observed in other tokamaks. For the deuterium retention, at least two retention processes (ion-implantation and co-deposition) were found on the dome region. The systematic dust collection gave the small amount of dust (∼7g: 0.2mg/s production) in the whole vessel of JT-60U. Deposition was observed at the remote area of the outer divertor region. [Copyright &y& Elsevier]

Abstract Targets of Reduced Activation Ferritic Martensitic (RAFM) steel F82H were exposed to low-energy (200 eV), high flux (about 1022 D/m2s) deuterium (D) plasma at 623–773 K to various D fluences in the range from 1 × 1025 to 2.5 × 1026 D/m2. The surface morphology of the plasma-exposed targets was examined with a field-emission scanning electron microscope. Cross-sectional observations of nano-structures formed on the F82H target surfaces were performed using a transmission electron microscope equipped with an energy dispersive X-ray spectrometer. It has been shown that nano-sized fiber-like layers are formed on the target surfaces under D plasma exposure. Micro-sized surface morphology pattern depends on the D fluence. As the D fluence increases, clusters of the fiber-like layers begin to be formed and organized into ordered structure. Highlights • F82H steel targets were exposed to 200 eV D plasma at elevated temperatures to various D fluences. • Nano-sized fiber-like layers were formed on the target surfaces. • As the D fluence increased, clusters of the fiber-like layers began to be formed and organized into ordered structure. [ABSTRACT FROM AUTHOR]

Surface topography of and deuterium (D) retention in reduced activation ferritic-martensitic Eurofer’97 and ferritic oxide dispersion strengthening ODS-16Cr steels have been studied after exposure at 600 K to low-energy (70 and 200 eV), high-flux (∼10 22 D/m 2 s) pure D and D-10%He plasmas with D fluence of 2 × 10 25 D/m 2 . The methods used were scanning electron microscopy, energy-scanning D( 3 He,p) 4 He nuclear reaction, and thermal desorption spectroscopy. As a result of the plasma exposures, nano-sized structures are formed on the steel surfaces. After exposure to pure D plasmas, a significant fraction of D is accumulated in the bulk, at depths larger than 8 μm. After exposures to D-He plasmas, D is retained mainly in the near-surface layers. In spite of the fact that the He fluence was lower than the D fluence, the He retention in the steels is one order of magnitude higher than the D retention. [ABSTRACT FROM AUTHOR]

Surface morphology changes and deuterium (D) retention in Toughened, Fine-Grained Recrystallized Tungsten (TFGR W) with TaC dispersoids (W-TaC) and pure tungsten exposed to D plasmas to a fluence of 10 26 D/m 2 s were studied as a function of the D ion flux (10 22 –10 24 D/m 2 s). As the flux increased from 10 22 D/m 2 s to 10 24 D/m 2 s, the numbers of blisters increased for both materials. However, smaller blisters were observed on W-TaC compared to pure W. In W-TaC, cracks beneath the surface along grain boundaries were observed, which were comparable to the blister sizes. The reason for the smaller blister sizes may arise from smaller grain sizes of W-TaC. In addition, reduction of the D retention in W-TaC was observed for higher flux exposures. D depth profiles indicate this reduction arises due to decrease in trapping in the bulk. [ABSTRACT FROM AUTHOR]

Surface topography and deuterium retention in polycrystalline hot-rolled W and W–5%Re have been examined after exposure to a low-energy (76 eV), high flux (around 10 22 D/m 2 s) deuterium plasma to an ion fluence of 10 26 D/m 2 at various temperatures. The methods used were confocal laser scanning microscopy and the D( 3 He, p) 4 He nuclear reaction at 3 He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 348 to 673 K, small blisters of size in the range from about 1 to about 15 μm, depending on the exposure temperature, are formed on the W and W–5%Re surfaces. In the W–5%Re, the deuterium retention demonstrates its maximum at exposure temperature of 463 K, while in the W this maximum is shifted to 523 K. A difference in the temperature dependence of the D retention for the W and W–5%Re is explained, as a rough approximation, by temperature dependences of the ductility of these materials. [ABSTRACT FROM AUTHOR]

Abstract: The tritium permeation behavior through F82H steel from approximately pure tritiated water vapor diluted with He was investigated and the as-received and the permeation-experimented sample surfaces were analyzed by means of SEM, EDX and XRD. Extremely slow but obvious tritium permeation through F82H steel could be observed. It is indicated from the surface analyses that an oxide layer composed of magnetite and hematite is formed during the permeation experiments. The permeabilities estimated using the partial pressure of molecular tritium under the extreme assumption that the redox reaction among tritiated water vapor, magnetite and hematite with forming molecular tritium reaches equilibrium indicate that apparent tritium permeation using tritiated water vapor for the inlet side would be proceeded by molecular tritium generated by the redox reaction, although it is expected that actual tritium permeation would be derived from dissociated tritium atom indicated in transition state during the redox reaction. [Copyright &y& Elsevier]

Abstract: We have investigated hydrogen isotopes retention in graphite tiles used as outboard first wall armors of JT-60U by means of SEM, TDS and SIMS. The SEM examination confirmed that these tiles are mostly eroded. Different from graphite divertor tiles, in which H retention was dominant owing to isotopic replacement during HH discharges preformed after DD discharges to remove tritium, deuterium is dominated in hydrogen isotopes retention of the outboard first wall tiles showing deeper penetration. This is attributed to the injection of high energy D and less isotopic replacement due to the lower temperature of the outboard first wall tiles. This type of hydrogen retention could not be avoided for the metallic wall. Although the amount of hydrogen isotopes retained per unit area in the outboard first wall tiles was much less than that for the divertor tiles, the integrated retention over the whole first wall surface could be very large. [Copyright &y& Elsevier]

Abstract: SlimCS is the conceptual design of a compact fusion DEMO plant assuming technologies foreseeable in 2020–2030s. For continuity of blanket technology from the Japanese ITER-TBM, the prime option of blanket is water-cooled solid breeder with Li2TiO3 (or Li4SiO4) and Be. A reduced-activation ferritic–martensitic steel (RAFM) and subcritical water are chosen as the structural material and coolant, respectively. The reactor has somewhat complex torus configuration with a sector-wide conducting plate slipped in between the replaceable (front) and permanent (back) blanket. In order to allow flexibility of maintenance in such a complex configuration, sector transport hot cell maintenance scheme is adopted. This paper describes characteristics of SlimCS with a focus on materials selection. [Copyright &y& Elsevier]

Low-Z impurity (7Be) on the JT-60U divertor tiles was analyzed to study the impurity behavior in the divertor region. The amount of the 7Be increased approximately one hundred times after B4C-tile installation in the outer divertor. The 7Be was probably produced by 10B(p,α)7Be nuclear reaction on the divertor tiles in the hydrogen experiment with ion cyclotron range of frequency heating. The 7Be was distributed asymmetrically in the poloidal and the toroidal direction. The highest 7Be concentration was found at the inner divertor whose boron content (B/(B+C) ∼ 20%) was lower than the B4C tiles (B/(B+C) ∼ 80%) of the outer divertor. This result may imply impurity transport from the outer divertor to the inner divertor. [Copyright &y& Elsevier]

Abstract: Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3×1020 H+Ds−1, and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world. [Copyright &y& Elsevier]

Abstract: Retention and depth profile of all hydrogen isotopes (hydrogen, deuterium and tritium) in the carbon tiles of outer divertor, outboard baffle plate and outboard first wall in JT-60U, which were mostly eroded, were measured by TDS, SIMS and TIPT, and discussed from view points of tile temperature, incident energy and flux of impinging H, D and T. The distribution of T in the analyzed tiles was compared with the impinging flux of tritium produced by DD reactions, which was calculated by the OFMC code. In this paper, H, D and T retention mechanisms and isotopic replacement effects are summarized. [Copyright &y& Elsevier]