Hisao Atsumi

Absorption and desorption of deuterium on graphite at elevated temperatures

Abstract The absorption and desorption behavior of deuterium has been studied on graphite exposed to a deuterium gas atmosphere at elevated temperatures. Thermal desorption measurements have been carried out at a constant heating rate of 10°C/min in vacuum. The solubility can be expressed by S(STP cm 3 /g) = 1.9 × 10 −4 p 1 2 (Pa)exp[19(kJ/mol)/RT] . Deuterium desorption curves on graphite seem to have three peaks at approximately 140°C, 480°C and 930°C. The second peak may be attributed to pore diffusion of deuterium expressed by D(cm2/s) =1800 exp[−121(kJ/mol)/RT]. The third peak may be attributed to bulk diffusion in graphite filler grains, and this can be expressed by D(cm2/s) = 1.69 exp[−251(kJ/mol)/RT].

Hydrogen solubility and diffusivity in neutron-irradiated graphite

Hydrogen solubility measurements and the analysis of absorption kinetics have been studied on graphite irradiated with neutrons at various fluences up to 5.4×10 24 n/m 2 . The absorption of hydrogen could be expressed as a diffusion-controlled process. The rate constant of hydrogen absorption was different from that for desorption. This difference may be ascribed to the effects of trapping sites in graphite. After neutron irradiation at 1.9×10 24 n/m 2 (~ 0.2 dpa), the hydrogen solubility was 20–50 times larger than that of unirradiated samples. The increase of hydrogen solubility was saturated above the damage level of ~ 0.3 dpa. The diffusivity of hydrogen was decreased by neutron irradiation up to 1.9×10 24 n/m 2 , and then increased above this fluence. This behavior can be ascribed to the production of the trapping sites for hydrogen, and the elongation of the distance between the basal planes by neutron irradiation.

Trapping and detrapping of hydrogen in carbon-based materials exposed to hydrogen gas

Abstract Measurements of hydrogen solubility have been performed for several unirradiated and neutron-irradiated graphite (and CFC) samples at temperatures between 700 and 1050°C under a ⋍ 10 kPa hydrogen atmosphere. The hydrogen dissolution process has been studied and it is discussed here. The values of hydrogen solubility vary substantially among the samples by up to a factor of about 16. A strong correlation has been observed between the values of hydrogen solubility and the degrees of graphitization determined by the X-ray diffraction technique. The relation can be extended even for the neutron-irradiated samples. Hydrogen dissolution into graphite can be explained by the trapping of hydrogen at defect sites (e.g. dangling carbon bonds) considering an equilibrium reaction between hydrogen molecules and the trapping sites. The migration of hydrogen in graphite is speculated to result from a sequence of detrapping and retrapping events with high-energy activation processes.

Radiation damage of graphite: degradation of material parameters and defect structures

Abstract On the basis of our previous work, we have discussed the relation between damage structure and the degradation of the material parameters of neutron-irradiated graphites. The defects produced in a basal plane and/or in between the basal planes (in-plane or two-dimensional defects), which are appreciable in the early stage of irradiation, seem to play a critically important role both in the reduction in the thermal conductivity and increase in the hydrogen retention. They also cause a dimensional change through increase in lattice spacing between the basal planes of graphite. Although the in-plane defects are rather easily annealed out, there seems to be no way to avoid their production under neutron irradiation, particularly at low temperatures. After heavy irradiation, the defects grow into three-dimensional clusters, probably accompanying some sp2-to-sp3 transition. They play an important role in volume expansion and result in complete loss of the layered structure of graphite (amorphization), which is very difficult to anneal. Considering the annealing behaviors of the thermal conductivity, lattice constant and electrical resistivity, we propose a new model based on the sp2-to-sp3 transition that can explain the observed effect for both damage and annealing processes without any contradiction.

Hydrogen behavior in carbon-based materials and its neutron irradiation effect

Abstract Hydrogen retention in graphites and CFCs (carbon fiber/carbon composites) has been studied with the crystallographic data obtained by the X-ray diffraction (XRD) technique. The amounts of retained hydrogen vary substantially among the samples by a factor of up to 16. After neutron irradiation at 1.9 × 1024 n/m2(∼ 0.2 dpa), the hydrogen retained becomes 20–50 times a larger than that of unirradiated samples. A strong correlation is observed between the values of hydrogen retention and the lattice constant c0. The size of crystallite also has a good correlation with the hydrogen retention. Hydrogen atoms will be trapped at dangling carbon bonds at edge surfaces of crystallite which are thermally stable even at high temperatures above 1000°C. Differences among the desorbed amount of hydrogen gas from graphite materials can be also explained well by this model.

Thermal desorption behavior of he implanted into graphite

Thermal desorption of He from various graphite samples irradiated with 20 keV He + ions has been studied. Thermal desorption behavior of He from Poco graphite was slightly different from that of Isograph-88 and above a dose of approximately 1018 ions/cm2 both isotropic graphites had almost identical desorption curves with peak temperatures of approximately 330°C. In the case of the graphitized paper Papyex, the amount of released He reached a maximum at a dose of 5 × 1017 ions/cm2 and the peak temperature on the thermal desorption curve became constant at approximately 200°C above a dose of 1 × 1018 ions/cm2. The activation energy (Ed) for He desorption from Poco graphite increased with the irradiation dose and was estimated to be 0.72 eV for a dose of 5 × 1016 ions/cm2, 0.90 eV for 5 × 1017 and 1 × 1018 ions/cm2, respectively.

Hydrogen absorption into neutron-irradiated graphite and estimation of the trapping effect

Bulk hydrogen retention and the analysis of absorption kinetics have been studied on graphite irradiated with neutrons at various conditions. Two kinds of hydrogen trapping sites may exist and be additionally produced during irradiation: interstitial cluster loop edge sites (trap 1) and carbon dangling bonds at edge surfaces of crystallites (trap 2). Neutron irradiation preferably creates trap 2 sites at lower fluences and trap 1 sites at a higher fluence. Trap 2 tends to be annealed out at high temperatures, although trap 1 is hardly decreased even at 1873 K. The activation energy of hydrogen diffusion is found to be increased from 1.04 to 1.60 eV by neutron irradiation.

Hydrogen absorption process into graphite and carbon materials

Abstract In order to estimate bulk hydrogen retention and recycling in graphite and carbon materials, molecular hydrogen absorption has been studied. Hydrogen absorption rates significantly depend on samples which arise from grain size, trap concentration and so on. Absorption rate constants differ between the cases of low and high pressures. Trapping has a strong influence, especially in the low pressure range. Oxidation of graphite reduces hydrogen retention and enhances the absorption rate. This suggests that oxygen in graphite does not behave as trapping sites for hydrogen. Activation energies for apparent diffusion for H 2 and D 2 are determined to be 153 and 158 kJ/mol. They are smaller than those energies determined from desorption measurements.

Dissolution and diffusion of helium in graphite

Abstract Thermal desorption measurements were performed on three isotropic graphite samples which were exposed to helium gas at 100–700° C under 7–130 kPa for 1 h. Helium atoms which were desorbed from the sample during ramp heating at 10° C/min in vacuum were identified with a quadrupole mass spectrometer. Helium was dissolved into graphite for a few ppm in atomic fraction. The thermal desorption curve for ISO-880U appears to consist of two peaks (450 and 620° C) after helium gas exposure at above 500° C. The helium desorption can be ascribed to the diffusion-controlled process where the helium atoms may leave through bulk surface and open pores.

Hydrogen retention in high-Z materials with various contents of carbon

In order to estimate hydrogen retention and recycling in high-Z plasma facing components, bulk hydrogen retention in molybdenum and tungsten co-existing with carbon has been studied. Hydrogen retention in powdered specimens of molybdenum and tungsten is considerably higher than that in sheet specimens due to surface impurities and defects. When molybdenum and tungsten are well carbonized, bulk hydrogen retention is drastically reduced. Coexisting carbon strongly suppresses hydrogen diffusion to reduce absorption rate into the specimen. Ion and neutron irradiation may produce free carbon to enhance the hydrogen retention in molybdenum and tungsten.