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A seven-membered carbon-ring-forming annulation occurred, upon treatment of heteroarylindoles with α,β-unsaturated carbonyl compounds in the presence of an indium catalyst. The reductive annulation was observed when octylsilane was used. With 1,3-dibromo-5,5-dimethylhydantoin instead of octylsilane, the oxidative annulation proceeded. These two reactions were completely controlled; this is the first achievement in the biheteroaryl-based seven-membered carbon-ring-forming annulation. Tetracyclic products obtained from the oxidative annulation underwent SEAr-based bromination regioselectively. The resulting Br–C bond was used for the palladium-catalyzed Suzuki–Miyaura cross-coupling reaction to extend the π-conjugated region. Reaction pathways were also discussed based on some mechanistic studies.

An efficient protocol for C–H condensation of hypervalent iodine compounds toward arenes in fluoroalcohols has been applied to the recyclable preparation of mesityl iodonium(III) salts. The electrophilicities of [hydroxy(tosyloxy)iodo]mesitylene (MesI(OH)OTs) and iodomesitylene diacetate (MesI(OAc)2) are suitably enhanced in 2,2,2-trifluoroethanol. A series of nucleophilic aromatic compounds react smoothly with MesI(OH)OTs and MesI(OAc)2 or in situ hypervalent iodine(III) species, generated from iodomesitylene, to provide the target mesityl iodonium(III) salts in good yields at room temperature with broad functional group tolerance. This C–H condensation strategy merits high para-regioselectivities during the diaryliodonium(III) salt formation, but the major limitation in the case of low-reactive aromatic substrates is byproduct formation resulting from the self-condensation of the nucleophilic mesitylene ring in MesI(OH)OTs and MesI(OAc)2.

Tritium permeation through and corrosion of structural materials are critical issues in fusion reactor liquid lithium-lead blanket concepts from the viewpoints of an efficient fuel cycle, higher operation rates, and radiological safety. In this study, lithium-lead compatibility of ceramic coatings has been investigated for the development of tritium permeation barriers with corrosion protection. Erbium oxide, yttrium oxide and zirconium oxide coatings were fabricated by a metal organic decomposition method on reduced activation ferritic/martensitic steel F82H substrates. Corrosion and delamination of the coatings were accelerated under a higher oxygen concentration. Under a lower oxygen concentration, zirconium oxide coatings had the best lithium-lead compatibility among three coating materials from surface and cross-sectional observations. However, the zirconium oxide coating after lithium-lead immersion at 550 °C for 500 h showed higher deuterium permeability in comparison to the sample without immersion. Formation of a chromium oxide layer on the surface of the substrate before fabricating the coatings drastically improved the lithium-lead compatibility of erbium oxide and yttrium oxide coatings. Degradation of the coatings was mainly caused by corrosion and delamination depending on immersion temperature, test duration, and impurity concentration.

Tritium permeation through structural materials is one of critical issues in liquid lithium-lead blanket concepts from the viewpoints of an efficient fuel cycle and radiological safety. Metal oxide coatings have been investigated as tritium permeation barrier and showed high permeation reduction factors. For the application to DEMO reactors, however, corrosion of the coatings by blanket materials is an unavoidable concern. This paper focuses on preparation of three metal oxide, erbium oxide, yttrium oxide, and zirconium oxide coatings by a liquid phase method and comparison of their properties in terms of hydrogen isotope permeability as well as lithium-lead compatibility. The deuterium permeation behavior of the erbium oxide and yttrium oxide coatings was similar, while the zirconium oxide showed a decrease of the permeation flux by further crystallization at lower temperature than the others. The zirconium oxide coating showed the best lithium-lead compatibility among three oxides at up to 600 °C. Deterioration of the coatings after static lithium-lead exposure would be caused by delamination and corrosion. Delamination of the coating would be prevented to control the coating-substrate interface. Corrosion of the coatings by formation of ternary oxides or reduction will be the main issue in lithium-lead compatibility at high temperatures.

Structures with pores having diameters of several nanometers can be used for distinguishing nucleotides in DNA. A 2-D material such as a thin membrane is expected to enhance precise molecule recognition. Here, we present the nanopore fabrication to thin 2-D membranes using He ion microscopy (HIM). We transferred three 2-D materials, chemical vapor deposition (CVD) graphene, CVD boron nitride, and peel-off MoS2, to the submicrometer pore of freestanding SiO2 membranes on a silicon substrate. The number of atomic layers of the 2-D materials was confirmed via Raman spectroscopy and X-ray photoelectron spectroscopy. We examined the reproducibility and accuracy of nanopore fabrication as a function of the focal position and beam irradiation intensity of the HIM system. Reproducible nanopores having diameters of less than 10 nm were realized by optimizing the intensity and focal position. We achieved 1.5-nm-diameter nanopores in a single layer of graphene using this HIM technique.

A method combined oxidation catalytic reactor and water bubbler is the most quantitative analysis method for gaseous tritium. We developed a tritium measurement system with high accuracy using an organic-based hydrophobic platinum catalyst previously. However, the previous measurement system had a shortcoming that the catalyst was combustible over 100 °C. To overcome the shortcoming, we developed an inorganic-based hydrophobic platinum catalyst (Pt/SiO 2 ) which was noncombustible. The developed Pt/SiO 2 catalyst was specially enhanced the hydrophobicity to prevent adsorption of water vapor on the surface of catalyst. To oxidize tritium efficiently, the particle diameter of 1mmϕ was selected for the catalyst and hydrogen was intentionally added at the inlet of the catalytic reactor to be the optimal concentration for oxidation. The performance tests showed that a modified quantitative tritium analysis system using by the hydrophobic Pt/SiO 2 catalyst provides tritium measurement with high accuracy more than 99.9%. We have confirmed the modified quantitative tritium analysis system is available in various conditions.

The effect of periodic irradiation on Deuterium (D) retention and blistering in Tungsten (W) was investigated. W samples were exposed to D plasma at a fixed fluence while varying the irradiation cycle number (1-shot, 2-shots and 3-shots). Exposure energy and flux were ∼50eV and ∼1 ×1022 D m−2 s−1, respectively. Sample temperatures were 537K and 643K. At 573K, D retention and blister density decreased with increasing number of irradiation cycle. In contrast at 643K, D retention showed no dependence on number of irradiation cycle. Therefore, sample temperature during irradiation is an important parameter in comparing the results of continuous and periodic irradiation, especially in studies involving extremely-high-flux (>1024 D m−2 s−1) irradiation and fluence dependency of D retention. Keywords: Tungsten, Hydrogen isotope retention, Blistering, Divertor, TDS

Recovery of hydrogen dissolved in Li-Pb eutectic alloy by mean of a bubbling tower is experimentally investigated. Mass-transfer coefficients to predict tritium recovery rate are experimentally determined when Ar and Ar+H2 gas bubbles are injected into Li-Pb through an I-shaped nozzle under the conditions of temperature 573–773 K and H2 partial pressure of 1 Pa–0.1 MPa. The results are fitted by an analytical equation based on diffusion and solution in Li-Pb. So that, the rate-determining step is hydrogen diffusion through a boundary layer formed in Li-Pb-gas interface and absorption and desorption are found to be almost reversible.

Thoughtful consideration of abnormal events such as fire is required to design and qualify a detritiation system (DS) of a nuclear fusion facility. Since conversion of tritium to tritiated vapor over catalyst is the key process of the DS, it is indispensable to evaluate the effect of excess moisture and hydrocarbons produced by combustion of cables on tritium conversion rate considering fire events. We conducted demonstration tests on tritium conversion under the following representative conditions: (I) leakage of tritium, (II) leakage of tritium plus moisture, and (III) leakage of tritium plus hydrocarbons. Detritiation behavior in the simulated room was assessed, and the amount of catalyst to fulfill the requirement on tritium conversion rate was evaluated. The dominant parameters for detritiation are the concentration of hydrogen in air and catalyst temperature. The tritium in the simulated room was decreased for condition (I) following ventilation theory. An initial reduction in conversion rat...

3,4-Diaminobenzenesulfonate (DBS) and PtII form a diradical complex (PtII–DBS) having cis/trans isomerism. Before separation of the isomers by using capillary electrophoresis (CE) using β-cyclodextrin (β-CD) derivatives as the selector, the inclusion behavior of PtII–DBS into β-CD and methyl-β-CD (MCD) was investigated using UV-vis and 1H NMR measurements to confirm the formation of a 1 : 1 inclusion complex at pH 10.0 in aqueous solutions. By employing CE using an MCD-containing electrophoresis buffer, the cis and trans isomers of PtII–DBS were separated. Single crystals of trans-PtII–DBS were obtained to permit the assignment of the resolved peaks. By analyzing the dependence of the electrophoretic mobility of the isomers on the concentration of CDs in the buffer, the thermodynamic inclusion constant was determined. For example, the inclusion complex of PtII–DBS into MCD afforded Kcis = 378 ± 2 and Ktrans = 330 ± 4 M−1 (at pH 10.0, 298 K). The difference in the Gibbs energy of inclusion, ΔΔG = −RT ln(Ktrans/Kcis), was only 0.34 kJ mol−1, suggesting that CE can discriminate between marginal differences in energy. The comparison in the electrophoretic mobilities of the inclusion complexes indicated that the cis isomer affords a value 21% greater than that observed for the trans isomer, suggesting that cis-PtII–DBS is tightly bound to the MCD cavity and forms a more compact complex than trans-PtII–DBS. Thus, CE using CDs as the tool not only resolves the isomers but also demonstrates the inclusion manner in solution.

To design a recombiner packed with hydrophobic catalyst for passive tritium oxidation, the development of manufacturing technology for hydrophobic platinum catalyst and the following analysis on reaction rate are essential. This presentation deals with the reaction rate for tritium oxidation over a hydrophobic platinum catalyst at temperature between room temperature and 473 K. Two kinds of hydrophobic platinum catalysts; supported with styrene-divinylbenzene and with silica were used for this test. The results indicate that 1) the particle size of platinum affects the reaction rate, 2) the order of reaction is 0.5 of hydrogen concentration at room temperature in case of a hydrogen content below 100 ppm. The rate-determining step approximation method is applied for the discussion on the reaction mechanism of tritium oxidation over platinum catalyst.

Various methods of tritium measurement have been applied depending on a chemical form of tritium. A method combined oxidation catalyst and water bubblers has been used as one of the most quantitative analysis methods for gaseous tritium. The major technical shortcoming with the conventional methods is the lowering of quantitative accuracy caused by adsorption of produced tritiated vapor on hydrophilic oxidation catalyst. Then, we developed an organic-based hydrophobic platinum catalyst. The catalyst enhanced its hydrophobicity to prevent absorption of water vapor on catalyst surface. In order to ensure to oxidize tritium, we additionally devised 1) the particle size of the catalyst made small to be around 1 mm ϕ and 2) intentional addition of hydrogen. As the result of experimental verification of the proposed method, we succeeded to achieve the high accuracy of more than 99.9% in the quantitative measurement of tritium. The developed tritium measurement system is applicable widely for various tritium measurements, estimation of tritium generation rate at TBM, tritium behavior in various materials and so on.

Yttrium oxide coatings were fabricated on reduced activation ferritic/martensitic steels by metal organic decomposition with a dip-coating technique, and their deuterium permeation behaviors were investigated. The microstructure of the coatings varied with heat-treatment temperature: amorphous at 670 oC (amorphous coating) and crystallized at 700 oC (crystallized coating). Deuterium permeation flux of the amorphous coating was lower than the uncoated steel by a factor of 5 at 500 oC, while that of the crystallized coating was lower by a factor of around 100 at 400‒550 oC. The permeation fluxes of both coatings were drastically decreased during the measurements at higher temperatures by a factor of up to 790 for the amorphous coating and 1000 for the crystallized one, indicating a microstructure modification occurred by an effect of test temperature with hydrogen flux. Temperature dependence of deuterium diffusivity in the coatings suggests that the decrease of the permeation flux has been derived from a decrease of the diffusivity. Characteristic permeation behaviors were observed with different annealing conditions; however, they can be interpreted using the permeation mechanism clarified in the previous erbium oxide coating studies.

The effects of tritium on the passivation behavior and passive layer formed in tritiated water circumstance for SS316L were investigated by means of an anodic polarization measurement technique and X-ray photoelectron spectroscopy, respectively. The results were compared with those for SS304, since it was predicted from a model of the tritium effects on corrosion suggested in the previous studies that SS316 would be less affective to tritiated water circumstance than SS304. As the results, the passivation inhibitory effect of tritium could not be observed for SS316L, while it was observed for SS304 and the other researched materials so far, as predicted. However, the thickness of the passive layer and the boundary between the passive layer and bulk of SS316L were found affected by tritium; thickened and gradated, respectively. From these results, it was concluded that SS316L would be more sustainable in tritiated water circumstance than SS304, although the corrosion of SS316L would be more or less enhanced in tritiated water circumstance.

It is known that the chemical forms of tritium released from a ceramic breeder blanket are hydrogen form and water form. To recover tritiated water vapor, adoption of dryer that is packed column of synthetic zeolite has been proposed. On the other hand, synthetic zeolite is often used as a support of precious metal catalyst. Such catalysts usually have a capability of hydrogen isotope exchange between gas and water vapor. If this catalyst is used to dryer, the dryer may obtain a preferable function for tritium recovery by isotopic exchange reaction. To assess such functions, reaction rate should be estimated. The results of water adsorption experiment on cation-exchanged mordenite-type zeolite suggested the possibility that state of adsorbed water varied by exchanged cation. So, in this work, precious metal catalyst based on cation-exchanged mordenite was prepared, and the reaction rate of chemical exchange between hydrogen and tritiated water was investigated under temperature range between 30 °C and 80 °C by the steady-state approximation. In the case of platinum on Na-mordenite, the reaction between gaseous hydrogen and tritiated water vapor was almost expressed as first-order reaction concerning tritiated water vapor concentration.

Iodonium(III) salts bearing uracil moieties have recently appeared in the literature, but their structural scope and utilization are limited because of their hygroscopic characteristics. In this study, we describe our detailed investigations for synthesizing a series of uracil iodonium(III) salts derived with various structural motifs and counterions. These new compounds have been utilized as attractive synthetic modules in constructing functionalized nucleobase and nucleosides.

Detritiation system of a nuclear fusion plant is mandatory to be designed and qualified taking carefully into consideration all the possible extraordinary situations in addition to that in a normal condition. We focused on the change in the efficiency of tritium oxidation of a catalytic reactor in an event of fire where the air accompanied with hydrocarbons, water vapor, and tritium is fed into a catalytic reactor at the same time. Our test results on the effect of these gases on the efficiency of tritium oxidation of the catalytic reactor indicated; (1) tritiated hydrocarbon produces significantly by reaction between tritium and hydrocarbons in a catalytic reactor; (2) there is little possibility of degradation in the detritiation performance because the tritiated hydrocarbons produced in the catalyst reactor are combusted; (3) there is no possibility of uncontrollable rise in the temperature of the catalytic reactor by heat of reactions; and (4) saturated water vapor could temporarily poison the...

Development of an efficient tritium recovery method is indispensable in order to compose a liquid blanket system of a D-T fusion reactor in the near future. Here, tritium recovery using a bubbling ...

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 1026 D/m2 s were studied as a function of the D ion flux (1022–1024 D/m2 s). As the flux increased from 1022 D/m2 s to 1024 D/m2 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.

The deuterium (D) permeation behaviors for ion-damaged tungsten (W) by 3 keV D2+ and 10 keV C+ were studied. The D permeability was obtained for un-damaged W at various temperatures. For both D2+ and C+ implanted W, the permeability was clearly reduced. But, for the D2+ implanted W, the permeability was recovered by heating at 1173 K and it was almost consistent with that for un-damaged W. In the case of C+ implanted W, the permeability was not recovered even if the sample was heated at 1173 K, indicating that the existence of carbon would prevent the recovery of permeation path in W. In addition, transmission electron microscope (TEM) observation showed the voids were grown by heating at 1173 K and not removed, showing the existence of damages would not largely influence on the hydrogen permeation behavior in W in the present study.

The effects of tritium on the corrosion behavior of chromium in 0.01 N sulfuric solution have been investigated in the present study. Electrochemical experiments have been carried our for pure chromium. At first, the concentration dependence of sulfuric acid solution on anodic polarization behavior of chromium was experimented, resulting in that 0.01 N one was found appropriate. The dependence of both dissolved oxygen and tritium concentration on anodic behavior of chromium were performed. It was found from that the self-passivation of chromium induced by dissolved oxygen was inhibited in tritiated solution resulting in the enhancement of the corrosion. As a consequence it is highly likely that the elution of chromium by highly oxidative radiolysis products would explain the passivation inhibitory effect of SUS304 stainless steel observed in tritiated solutions.

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 1022 D/m2 s) deuterium plasma to an ion fluence of 1026 D/m2 at various temperatures. The methods used were confocal laser scanning microscopy and the D(3He, p)4He nuclear reaction at 3He 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.

To investigate the behavior of hydrogen on a water–metal boundary, a series of experiments have been performed that studied tritium permeation into a pressurized water jacket through pure iron piping, which contained approximately 1 kPa of pure tritium gas at 423 K, while monitoring the chemical forms of tritium. Additionally, the behavior of deuterium, which was generated on the heavy water–metal boundary and transferred to the metal, was also investigated using a pressure-proof heavy water vessel. Actual deuterium transfer through various metal piping, such as Fe, Ni, SS304, etc., was detected clearly by QMS at 423–573 K. Moreover, using the above heavy water system, we have succeeded in detecting hydrogen isotopes simultaneously transferring from and to the metal surface by introducing hydrogen gas to the Ni piping after deuterium permeation from the heavy water side had stabilized.

Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its property is changed easily by exchanging its cations. Synthetic mordenite is one of zeolites having fairly large silica/alumina ratio. There are many reports about hydrogen adsorption properties of cation-exchanged mordenite so far. And, the present authors also have reported that cation-exchanged mordenite with Ca ion (Ca–MOR) indicated fairly large hydrogen adsorption capacity at 77 K in comparison with Molecular Sieves 5A (MS5A). So, in this work, hydrogen adsorption properties of cation-exchanged mordenite with transition metal ion were investigated mainly. The cation-exchanged mordenite with Ag ion (Ag–MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77 K in comparison with Ca–MOR. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag–MOR is adopted.

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 + D s−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.

Retentions of Hydrogen (H) and deuterium (D) in the side surfaces (gaps between tiles) of the carbon tiles used as first wall in JT-60U were measured by the thermal desorption spectroscopy. In the ...

A passive reactor for tritium oxidation at room temperature has been widely studied in nuclear engineering especially for a detritiation system (DS) of a tritium process facility taking possible extraordinary situation severely into consideration. We have focused on bacterial oxidation of tritium by hydrogen-oxidizing bacteria in natural soil to realize the passive oxidation reactor. The purpose of this study was to examine the feasibility of a bioreactor with hydrogen-oxidizing bacteria in soil from a point of view of engineering. The efficiency of the bioreactor was evaluated by kinetics. The bioreactor packed with natural soil shows a relative high conversion rate of tritium under the saturated moisture condition at room temperature, which is obviously superior to that of a Pt/Al2O3 catalyst generally used for tritium oxidation in the existing tritium handling facilities. The order of reaction for tritium oxidation with soil was the pseudo-first order as assessed with Michaelis-Menten kinetics model. Our engineering suggestion to increase the reaction rate is the intentional addition of hydrogen at a small concentration in the feed gas on condition that the oxidation of tritium with soil is expressed by the Michaelis-Menten kinetics model.

In a fusion reactor, the prediction of tritium release behavior from breeder blanket is important to design the tritium recovery system, but the amount of tritium generated is necessary information to do that. Hence, tritium generation and recovery studies on lithium ceramics packed bed have been started by using fusion neutron source (FNS) in Japan Atomic Energy Agency (JAEA). Lithium titanate (Li 2 TiO 3 ) was selected as tritium breeding material, and its packed bed was enclosed by the beryllium blocks, and was kept at certain temperature during fusion neutron irradiation. During irradiation, the packed bed was purged with the sweep gas continuously, and tritium released was trapped in each gas absorber selectively by chemical form. In this work, the effect of sweep gas species on tritium release behavior was investigated. In the case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in the case of sweep by helium without water vapor, tritium in gaseous form was released first, and release of tritium in water form was delayed from gaseous tritium and was gradually increased.

In order to investigate the behavior of hydrogen isotope on the water–metal boundary, deuterium permeation experiments from heavy water vessel through various metal piping, such as pure iron (Fe), nickel (Ni), stainless steel (SS304), and pure iron with 10 μm gold plating, were performed at 573 K and at 15 MPa. During the experiment, surfaces of metal piping except gold plating one were oxidized at the heavy water boundary and then deuterium would generate by the oxidation reactions. This deuterium could be detected by mass spectrometer, which monitored the inside gases of the piping under continuous evacuation. The result showed clearly that the deuterium permeated through Fe, Ni, and SS304 piping was detected as mainly deuterium gas (D2) under continuous evacuation, though that through gold plating one could not be detected effectively. The D2 permeation rate through Fe, Ni, and SS304 piping reached equilibrium conditions with oxide generation at D2O–metal boundary, although concluding the transfer mechanism will require further testing and modeling activities.

The effects of tritiated water on the corrosion behavior of SUS304 stainless steel was studied using Tafel extrapolation method, one of electrochemical techniques with changing tritium concentratio...

Transport properties of tritiated water vapor (HTO) in the epoxy paint such as adsorption, desorption, diffusion and dissolution has been evaluated by investigating the HTO exposure and removal behavior from the epoxy paint in order to generate a data base on tritium behavior in tritium-confinement facilities such as the Hot Cell and the tritium plant building in ITER. Two types of experiments were carried out; one was the HTO exposure to the epoxy paint, and the other was the detritiation curves from the epoxy paint after the HTO exposure. Stainless steel vessel chambers with the epoxy painted inside surfaces were first exposed to an air flow containing HTO vapor (740 Bq/cm3) for 1 week, 2 weeks and 2 months. After these exposures, detritiation of these chambers with an air flow was carried out. It was found that the interaction between the surface of the epoxy paint and the HTO such as adsorption and desorption is reached the steady state under these conditions. Based on experimental detritiatio...

The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. It was found that the permeation rate of tritium was almost the same as that with no water vapor, and the major chemical form of tritium was HTO. It was also found that the permeation rate was decreased to ∼1/10 of the above value by plated with gold, and the major chemical form was HT. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). A set of data for an advanced chemical exchange column has been obtained as R&D on the tritiated water processing. The packing and catalyst were uniformly mixed into the advanced column. It has been proved that there was possibility that the advanced column has a large separation factor in comparison with the ordinary columns, where the packing and catalyst were alternately packed. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column. The ion conductivity of the main material of the cell (Nafion) was hardly decreased even in the case of 1600 kGy irradiation by gamma ray.

The ion fluence dependence of deuterium retention in SS-316 during oxidation at a temperature of 673 K was studied to evaluate the dynamics of deuterium retention in the oxide layer of SS-316. The correlation between the chemical state of stainless steel and deuterium retention was evaluated using XPS and TDS. It was found that the major deuterium desorption temperatures were located at around 660 K and 935 K, which correspond to the desorption of deuterium trapped as hydroxide. The deuterium retention increased with increasing deuterium ion fluence, since the deuterium retention as hydroxide increased significantly. However, retention saturated at an ion fluence of ∼2.5 × 1021 D+ m−2. The XPS result showed that FeOOD was formed on the surface, although pure Fe also remained in the oxide layer. These facts indicate the nature of the oxide layer have a key role in deuterium trapping behavior.

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.

A fusion reactor requires high levels of safety and public acceptability, so safeconfinement of tritium is one of the key issues. Tritium must be well controlled with no excessive release to enviro...

Concentration of tritium in highly tritiated water was measured by exposing imaging plates (IPs) to water vapor. Tritium penetrated into photostimulated luminescence (PSL) phosphor through polyethy...

A fusion reactor requires high levels of safety and public acceptability, so safe confinement of tritium is a key issues. In order to understand tritium behavior in the concrete reactor components, sorption and desorption experiments were carried out as a function of the exposure time for sorption and the kind of purge gas for desorption. The samples used were cement paste, mortar and concrete. These samples were exposed to 740–1110 Bq/cm 3 of HTO vapor in 1 kPa of H 2 O vapor at room temperature. Desorption experiments were carried out using N 2 purging gas with water vapor by the thermal desorption method. The concrete materials became almost saturated with HTO within about 1–6 months. Though the chemical form of tritium desorbed from the sample was mostly HTO, HT and other compounds were also detected above 673 K.

In order to investigate the behavior of hydrogen isotope on the water-metal boundary, a series of deuterium permeation experiment from heavy water vessel through pure iron piping was performed as a...

The trapping and release mechanisms of hydrogen isotopes for the stainless steel (SS) oxidized at various temperatures were investigated. The oxide layer was mainly consisted of iron oxides (FexOy)...

Since exotic corrosion of stainless steels in tritiated water can be expected, the anodic polarization of a SUS304 stainless steel sample in approximately 5 wt% sulfuric acid solution was performed at various concentrations of tritium and dissolved oxygen (hereafter DO) in the electrolyte. The inhibitory effect of tritium on the passivation could be observed with DO even at a tritium concentration in the electrolyte of as low as 2.2 kBq cm−3. This effect became more pronounced as the tritium concentration increased. It was suggested that the inhibitory reaction depending on tritium concentration would compete with the self-passivation depending on the DO concentration (hereafter [DO]), since it was found that there is a threshold [DO] for self-passivation at each tritium concentration.

The deteriorations of polymeric materials for a SPE-type high-level tritiated water electrolyzer composed of the Water Detritiation System (WDS) against sulfonic acid environment and radiation environment were discussed. A long-term durability of VITON, AFLAS, denaturated polyphenylene ether, and Kapton polyimide immersed in a sulfonic acid was demonstrated. Negligible degradation in percent elongation at break of these polymeric materials was observed up to the immersing period of 2 years. The detectable radiation deterioration in ionic conductivity of Nafion® N117CS ion exchange membrane irradiated with electron beams up to the integrated dose of 1500 kGy was measured. The ionic conductivity of Nafion® N117CS ion exchange membrane irradiated at more than 1000 kGy was slightly deteriorated. As for the elastomers for its use as a seal, the radiation degradation in hardness of VITON, AFLAS was investigated. Considerable degradation in softness of these rubbers for its use as a seal was not observed up to the integrated dose of 1500 kGy. The water uptake of rubbers was generally increased as the integrated dose was increased. However, irradiated VITON rubbers had constant water uptake up to the integrated dose of 1500 kGy.

We have developed the short time exposure (STE) test using a rabbit corneal cell line (SIRC cells) as an alternative eye irritation test. The STE test uses relative viability as the endpoint after cells are exposed to the test material at 5%, 0.5%, and 0.05% concentrations for 5 minutes. In this interlaboratory study, 2 laboratories conducted the test using 70 raw materials in order to evaluate transferability, between-laboratory reproducibility, and predictive capacity of the STE test as an alternative eye irritation test. Transferability was assessed using saline as a negative control and 0.01% sodium lauryl sulfate as a positive control. The relative viabilities obtained for the 2 laboratories were almost the same. Therefore, transferability was considered to be excellent. The 2 laboratories showed similar relative viabilities for all 70 raw materials at each test concentration. The correspondence rates of the eye irritation categories (irritants and nonirritants) were over 97% for each concentration tested, exhibiting high between-laboratory reproducibility. The correspondence rates for predicting eye irritation potential of undiluted raw materials and a diluted solution (10%) were over 85% at each laboratory for the 5% and 0.05% test concentrations in the STE. Lastly, the correspondence rate for the rank classification by the STE test prediction model at each laboratory was over 72%, and the correspondence rate became almost 90% when acids, amines, and alcohols were excluded from the analysis. From the above data, excellent transferability, high between-laboratory reproducibility, and high predictive capacity of the STE test were observed in the interlaboratory study by 2 laboratories.

Temperature dependence of oxide layer formation on hydrogen isotope retention in stainless steel type 316 was studied by TDS and XPS. The shape of TDS spectrum was clearly changed by the oxide formation temperature. The chemical states of iron, chromium and oxygen were also evaluated by XPS. The surface oxide layer was composed of iron and oxygen and the contribution of chromium was quite low. The ratio of oxide layer on stainless steel increased as increasing the annealing temperature. The deuterium retention trapped by the oxide layer, which corresponded to the desorption temperature of 600-800 K, was governed by the ratio of oxide layer, especially iron hydroxide. All of the iron was not oxidized and the saturation ratio of iron oxide to pure iron existed in the stainless steel. It was concluded that the saturation of deuterium retention trapped by the oxide layer was controlled by the amount of iron oxide in the oxide layer.

In order to investigate the function of water molecule for tritium transport behavior on the water-metal boundary, a series of experiments of tritium permeation into humid atmosphere was performed through pure iron piping with different surfaces of oxide etc., which contained about 1 kPa of pure tritium gas at 423 K. Chemical forms of tritium permeated into water were monitored continuously under purging outer jacket by

Synthetic zeolite is the candidate material of the separation column of the gas chromatograph for the hydrogen isotope analysis. Mordenite (MOR) is one of the zeolite, and has been reported that th...

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.

In an ITER facility, about 3 kg of tritium will be stored in about 50 hydride beds. The safe design and operation of tritium storage beds will be one of the most important points for enhancing overall safety in the facility. In the Tritium Process Laboratory at the Japan Atomic Energy Agency, many tritium storage beds with zirconium–cobalt (ZrCo) and uranium (U) have been used with/without self-accountancy measures, and 20 years of safe handling experience has been accumulated. From this experience, the key issues to be considered for the safety design are the effects of tritium decay, such as decay heat transfer and 3He behavior, which can cause an increase in temperature and pressure in the tritide vessel, along with the normal protection against over-temperature, over-pressure and leaks in the metal-hydride bed. Concerning the safety operation, the key issues are the hydrogenation–dehydrogenation cycle sequence under the requirements of the storage system and the emergency action, such as rapid hydrogen recovery and loss of normal cooling function.

To clarify the tritium permeation behavior, tritium distribution in iron oxidized in high temperature water was observed with tritium micro autoradiography. It was found that tritium was distributed homogeneously in the iron metal. However the oxide surface (magnetite) was found to contain a very low concentration of tritium. The inner layer of oxide could strongly effect the tritium permeation. From a comparison with the permeation experiment that had been reported in Ref. 1, it was suggested that tritium would mainly diffuse other path except the oxide lattice. According to the chemical form of tritium, which was released from iron surface into water, two assumptions were suggested. One is based on the different combination of tritium on the water-surface interface. The other is based on the oxidation mechanism. (authors)

To construct a fusion reactor with high safety and acceptability, it is necessary to establish and to ensure tritium safe handling technology. Tritium should be well-controlled not to be released to the environment excessively and to prevent workers from excess exposure. It is especially important to grasp tritium behavior in the final confinement area, such as the room and/or building. In order to obtain data for actual tritium behavior in a room and/or building, a series of intentional Tritium Release Experiments (TREs) were planned and carried out within a radiologically controlled area (main cell) at Tritium System Test Assembly (TSTA) in Los Alamos National Laboratory (LANL) under US-JAPAN collaboration program. These experiments were carried out three times. In these experiments, influence of a difference in the tritium release point and the amount of hydrogen isotope for the initial tritium behavior in the room were suggested. Tritium was released into the main cell at TSTA/LANL. The released tritium reached a uniform concentration about 30 ∼ 40 minutes in all the experiments. The influence of the release point and the amount of hydrogen isotope were not found to be important in these experiments. The experimental results for the initial tritium behavior in the room were also simulated well by the modified three-dimensional eddy flow analysis code FLOW-3D.