Thermal desorption of deuterium from isotropic graphites and a CFC material priorly charged with deuterium gas has been investigated to obtain quantitative information on hydrogen recycling and tritium inventory in fusion experimental devices and future reactors. The spectra of thermal desorption spectrometry (TDS) appear to have five peaks, which are defined as peaks 1–5 in order of the increase of temperature. Diffusion coefficients of deuterium for the absorption process are evaluated for the samples charged with deuterium gas for various absorption times. Absorption pressure dependence of TDS spectra shows a unique change in the temperatures of peak 2. As for the absorption temperature dependence of TDS spectra, peak 4 has a drastic change in the shape of spectra and also in the peak temperatures. The mechanism of desorption for peak 5 will be associated to a detrapping process with a dispersion of energies in the range of 4.0–4.9 eV.

Thermal desorption spectrometry (TDS) has been investigated to obtain fundamental information of tritium behavior in graphite and carbon materials especially at high temperatures. 29 brands of graphite, HOPG, glassy carbon and CFC materials charged with deuterium gas are tested up to the temperature of 1735 K with a heating rate of 0.1 K/s. TDS spectra have five peaks at 600-700 K, around 900 K, 1200 K, 1300-1450 K and 1600-1650 K The amounts of released deuterium have been compared with crystallographic parameters derived from XRD analysis. Reduction of tritium retention and inventories are discussed.

Thermal desorption behavior of deuterium (D-2) from isotropic graphites and a carbon fiber carbon composite (CFC) charged with D-2 gas has been investigated to obtain information concerning hydrogen recycling and tritium inventory in fusion experimental devices as well as a futuristic fusion reactor. After thermal desorption experiments were conducted at temperatures up to 1740 K, a desorption peak at approximately 1600 K (peak 4) was discovered. This is in addition to the previously known peak at approximately 1300 K (peak 3). Peak 3 can be attributed to the release of deuterium controlled by the diffusion process in a graphite filler grain and peak 4 can be attributed to the detrapping of deuterium released from an interstitial cluster loop edge site. Activation energies of peaks 3 and 4 are estimated to be 3.48 and 6.93 eV, respectively. (C) 2013 Elsevier B. V. All rights reserved.

Thermal desorption spectrometry (TDS) has been investigated to obtain fundamental information of hydrogen behavior in graphite and CFC especially at high temperatures. Thirteen brands of graphite and CFC materials charged with deuterium gas are tested up to the temperature of 1720 K with a heating rate of 0.1 K/s. TDS spectra have at least four peaks at 600-700 K, around 900 K, 1300-1450 K and 1600-1650 K. The change of TDS spectra is measured for the samples, which are charged with deuterium at 1273 K under a different pressure in the range of 83 Pa to 79 kPa. Physical and chemical states of deuterium in graphite and mechanisms of desorption are discussed. (C) 2013 Elsevier B.V. All rights reserved.

実用化以上のNbを添加した様々な高Nb含有Zr合金を真空中で熱処理した後、空気中に保存し、熱処理温度、時間による水素吸収特性及び窒化膜厚変化による水素拡散障壁の効果を検討した。

Bulk hydrogen retention and the mechanisms of hydrogen transport in graphite and carbon materials are summarized in the first half of the paper. From the various data on hydrogen retention, diffusion coefficients, nature of trapping sites and annealing effects, simulation of bulk hydrogen retention is performed. Neutron irradiation tends to increase hydrogen retention approximately 100 times larger than that of unirradiated graphite. (C) 2010 Elsevier B.V. All rights reserved.

Bulk hydrogen retention and the analysis of absorption kinetics have been studied on graphite samples irradiated with neutrons at various fluences. There are two kinds of hydrogen trapping sites: interstitial cluster loop edge sites (trap 1) and carbon dangling bonds at edge surfaces of a crystallite (trap 2). Neutron irradiation preferably creates trap 2 sites at lower fluences and trap I sites at higher fluences. The dissociation enthalpy becomes 2-3 times higher for neutron-irradiated samples due to the creation of high energy trapping sites. Absorption kinetics, such as diffusion coefficients and absorption rate constants are strongly affected by neutron irradiation and annealing after irradiation. (C) 2009 Elsevier B.V. All rights reserved.

Bulk hydrogen retention and the analysis of absorption kinetics have been studied on three brands of graphite irradiated with neutrons at various fluences. 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 above 0.017 dpa. Trap 2 tends to be annealed out at high temperatures, although trap 1 is hardly decreased even at 1873 K. Diffusion coefficients of hydrogen are reduced for 1-2 orders of magnitude after neutron irradiation at 0.047 dpa, although the activation energy of hydrogen diffusion is kept nearly the same level to unirradiated samples. (C) 2008 Elsevier B.V. All rights reserved.

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.

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. © 2007 The Royal Swedish Academy of Sciences.

Hydrogen absorption and transport in graphite materials have been studied to obtain fundamental information for a fusion reactor application and for hydrogen storage materials. Two kinds of hydrogen trapping sites should exist. One will be carbon dangling bonds located at the edge surface of a crystallite with an adsorption enthalpy of 2.6 eV, the other will be a solitary carbon dangling bond introduced by ion or neutron irradiation, such as an interstitial cluster loop edge, with an enthalpy of 4.4 eV. The results are compared with thermal desorption spectra of deuterium from graphites which were gas charged, ion irradiated and nano-structured by ball milling. These spectra can be well represented with three kinds of desorption process with activation energies of 1.3, 2.6 and 4.4 eV. (C) 2003 Elsevier B.V. All rights reserved.

Bulk hydrogen retention and hydrogen diffusion in graphite and carbon materials have been studied to estimate hydrogen recycling and tritium inventory under a fusion reactor environment. Two kinds of hydrogen trapping sites may exist. The first will be one of lined carbon dangling bonds located at the edge surface of a crystallite with an adsorption enthalpy of 2.6 eV, the second will be a solitary carbon dangling bond, such as an interstitial cluster loop edge, with an enthalpy of 4.4 eV. The correlation between hydrogen retention and the microstructure should refer to the edge surface area of a crystallite for an unirradiated sample and to lattice spacing along the c axis for an irradiated sample. The diffusion process is the rate-determining step for hydrogen absorption into graphite, and detrapping dominates the hydrogen desorption process due to the high trapping energy. © 2003 Elsevier Science B.V. All rights reserved.

Bulk hydrogen retention and hydrogen diffusion in graphite and carbon materials have been studied to estimate hydrogen recycling and tritium inventory in a fusion reactor environment. Hydrogen may permeate into a filler grain in the form of a hydrogen molecule. diffuse through crystallite boundaries and finally be trapped as hydrogen atoms at the edge surface of a crystallite. In the estimation of hydrogen transport, the activation energy of hydrogen diffusion can be determined from absorption experiments. and the activation energy of detrapping can be obtained from desorption experiments. The activation energies of hydrogen trapping are 2.6eV in ordinary graphite and 4.4eV for irradiated or mechanically milled graphite samples.

Molecular hydrogen absorption into graphite has been studied in order to obtain information on the 'true' hydrogen diffusion coefficient in graphite, oxygen effect and the mechanism of hydrogen trapping, which should be important issues on estimating hydrogen retention and recycling in plasma facing graphite and CFC. Hydrogen may permeate into a filler grain in the form of hydrogen molecules, diffuse through crystallite boundaries, and may finally be trapped as hydrogen atoms at the edge surface of a crystallite. The diffusion coefficient can be given as D.(m(2)/s) 3.3 x 10(-10)(-1.3 eV/kT), when the trapping effect does not exist. The simulation with mass balance equations can reproduce change of apparent diffusion coefficients. (C) 2002 Elsevier Science B.V. All rights reserved.

The benefits of using CVI SiC/graphite fiber composites as low tritium retaining, high thermal conductivity composites for fusion applications are presented. Three-dimensional woven composites have been chemically vapor infiltrated with SiC and their thermophysical properties measured. One material used an intermediate grade graphite-fiber in all directions (Amoco P55) while a second material used very high thermal conductive fiber (Amoco K-1100) in the high fiber density direction. The overall void was less than 20%. Strength as measured by four-point bending Was comparable to those of SiC/SiC composite. The room temperature thermal conductivity in the high conductivity direction was impressive for both materials, with values >70 W/mK for the P-55 and >420 W/m-K for the K-1100 variant. The thermal conductivity was measured as a function of temperature and exceeds the highest thermal conductivity of CVD SiC currently available at fusion relevant temperatures (>600 degreesC). Limited data on the irradiation-induced degradation in thermal conductivity is consistent with carbon fiber composite literature. (C) 2002 Elsevier Science B.V. All rights reserved.

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. (C) 2000 Elsevier Science B.V. All rights reserved.

Structural changes of various RFe2 intermetallic compounds and their conditions for hydrogen-induced amorphization (HIA) are studied with the techniques of X-ray diffraction, transmission electron microscope and pressure-composition isotherms. CeFe2 are amorphized even at 293 K under a hydrogen pressure of 0.1 MPa. SmFe2, GdFe2 and DyFe2 are amorphized in narrow temperature ranges at around 530 K. ErFe2 are not amorphized at such a low hydrogen pressure of 0.1 MPa. For GdFe2, HIA can be obtained at 533 K above an equilibrium hydrogen pressure of 2 kPa. The composition of amorphous hydrides has the hydrogen content above GdFe2H2.2.

In order to estimate hydrogen retention and recycling in high-Z plasma facing components, bulk hydrogen retention in molybdenum and tungsten cct-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. (C) 1998 Elsevier Science B.V. All rights reserved.

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 larger than that of unirrudiuted 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. Copyright © 1996 Elsevier Science B.V. All rights reserved.

For the DyFe2 Laves phase intermetallic compound, the nature of hydrogen-induced amorphization, structural changes and the dynamic behavior of hydrogen atoms were studied by means of X-ray diffraction, magnetic after-effect and thermal desorption spectrometry. DyFe2 can be amorphized by hydrogen charging within the narrow temperature range between 523 K and 593 K, even at a low hydrogen pressure of 0.1 MPa. At lower temperatures, DyFe2 forms a crystalline hydride, whereas it decomposes into alpha-Fe and DyH2 at higher temperatures. From the relaxation spectra obtained in magnetic after-effect measurements, it is found that hydrogen atoms in the crystalline hydride occupy the tetrahedral interstitial sites consisting of DyFe3 and Dy2Fe2 with activation energies of short-range diffusion determined as 0.24 eV and 0.35 eV respectively. For the amorphous hydride, hydrogen atoms additionally occupy Dy3Fe sites with an activation energy of 0.43 eV, and the distribution of activation energies becomes broader.

The amorphization kinetics of the Fe2Ce Laves phase were studied, utilizing X-ray diffraction (XRD), differential scanning calorimetry (DSC) and thermal H desorption measurements. Additionally, the local 1-1 diffusion process in this ferromagnetic compound was investigated by means of magnetic susceptibility and magnetic after-effect (MAE) measurements. Fe2Ce can easily be amorphized (p(H2) = 5 x 10(3) to 1 x 10(5) Pa, T < 400 K); irreversible hydrogen-induced amorphization takes place in a narrow phase boundary which migrates from the surface into the sample. Furthermore, cyclic H-charging of amorphous Fe2Ce below its crystallization temperature reveals a severe degradation in H-storage capacity from 53 at.% to 18 at.% after only three cycles. This degradation is attributed to a hydrogen-induced phase separation into crystalline CeH2 and alpha-Fe which is enhanced by the high mobility of Fe atoms.

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. © 1995.

Measurements of hydrogen solubility have been performed for several unirradiated and neutron-irradiated graphite (and CFC) samples at temperatures between 973 and 1323K under a similar to 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 up to about a factor of 16. A strong correlation has been observed between the values of hydrogen solubility and the degrees of graphitization determined by X-ray diffraction technique. The relation can be extended even for the neutron irradiated samples. Hydrogen dissolution into graphite can be explained with 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.

The thermal properties of zinc · chromium spinel-chromium system were measured by flash method. The heat capacity of composite compacts changes approximately linearly with Cr content, but the thermal diffusivity was increased when a proper quantity of metallic chromium was added to the ZnCr₂O₄-Cr solidification system. A remarkable increase in thermal conductivity was observed when over 50 mass percent of Cr content. This corresponded to the proportion of ZnCr₂O₄ spinel and metallic chromium of composite compacts. And this was attributed to the electron conductivity increase due to the increase of metallic chromium in ZnCr₂O₄-Cr solidification system. The thermal expansion coefficient was increased with an increase of metallic chromium content. This was explained by the X-ray diffraction evidence that formation of continuous phase of metallic chromium in composite compacts.

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 {reversed tilde equals} 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. © 1994.

Hydrogen solubility measurements and the analysis of absorption kinetics have been studied on graphite irradiated with neutrons at various fluences up to 5.4×1024 n/m2. 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 1.9×1024 n/m2 (~ 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 1.9×1024 n/m2, 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. © 1992, All rights reserved.

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. © 1991.

"Overall Characterizations of Graphites as Fusion First Wall Material and Evaluation of the Stability against Plasmas" Interim Report by Fusion First Wall Material Research Group. Nuclear Fusion Reasearch Project, The Ministry of Education, Science and Culture, Japan

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. © 1988.

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 cm3/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]. © 1988.

核融合次期装置のプラズマ第一壁候補材として注目されている高融点金属材料(タングステン、モリブデン)中への水素吸収挙動を調べた。昨年度のタングステンを主とした測定品引き続き、本年度は、モリブデン、およびモリブデン炭化物についても、炭素含有率を変化合せた金属炭化物中への水素吸収と吸収速度の測定を行った。炭化物は、これらの金属粉末とカーボンブラック粉末を混合して1500℃、2時間の真空焼鈍を行うことで作製し、粉末X線回折によりその形成と残留炭素量を定量した。モリブデンは、この条件で、概ね炭化物となることが確かめられた。タングステンの場合、未反応炭素が比較的高い水素リテンションの原因となるとともに、酸素不純物が水素を強く捕捉することがわかった。一方、モリブデンの場合には、残留炭素が少ないにも関わらず、高い水素リテンションを示し、炭化物そのものが水素を捕捉するものと考えられ、酸素不純物の影響は、むしろ少ないことがわかった。このように、この2つの高融点金属および炭化物は、全く異なった水素吸収挙動を示した。これらの成果は、第8回核融合炉材料国際会議、日本原子力学会1998年春の年会、Journal of Nuclear Materialsに発表し、今後も引き続いて報告する予定である。

本研究では、核融合実験装置で使用されるプラズマ対向壁材料として、炭素系材料と高融合金属(タングステン、モリブデン)をとりあげ、核融合炉水素リサイクリング挙動解明の基礎データとなる高温での水素保持(リテンション)の測定を行った。また、この挙動に関する材料の構造との相関を調べるにあたり、粉末X線回析を中心とした材料のキャラクタリゼーションを行った、水素保持の測定では、1000℃程度の高温で、気相の水素から材料中への吸収を調べ、炭素材料では、その銘柄(製造会社)毎に大きな差違があることを見出した。同時に行った日本原子力研究所材料試験炉(JMTR)での中性子照射では、これらの試料で照射後に最大50倍の水素保持量を示すなど大きな変化が起こることがわかった。X線回析によれば、中性子照射した炭素材料では、照射による結晶のc軸方向への伸びに加えて結晶子サイズ(主としてa軸平行方向)が著しく減少することが確認され、結晶子の外周部面積の増大と水素保持量に良い相関があることがつきとめられた。以上のことより、炭素系材料における材料構造と水素保持量の関係が明らかとなり、将来のD-T反応で生じる14MeV中性子によって水素保持量、リサイクリング率が増大することが予想され、これらのシミュレーションを行うことができた。また、高融点金属の場合、単体の金属では水素保持量が少ないのに対して、炭素による表面汚染があれば保持量が増大することがわかり、このような炭素の効果を詳しく調べておく必要性のあることが改めて確認された。これらの成果は、隔年で開催されている核融合炉材料国際会議等において発表を行っている。

1.序 大型核融合実験装置で使用されている炭素系材料は,耐熱性,耐熱衝撃性に優れる反面,水素の吸蔵量が大きく,現状ではヘリウム放電等で水素量を下げてリサイクリングを制御している.しかし,今後の長時間放電に向けて,この水素リサイクリング過程についての知見を得ることが重要となっている.本研究では,まず高温における炭素材料中の水素リテンションを知ることを目的として,水素溶解測定を行い,水素溶解の機構とリテンション低減への提案を行なった. 2.方法 微量水素溶解度測定装置を用いて,20銘柄の試料について,水素溶解度,溶解速度の測定を行なった.この溶解に及ぼす試料の特性については,粉末X線回析を中心に測定を行い,溶解挙動に及ぼす中性子照射効果をJMTR照射試料について試験した. 3.結果 水素溶解度は,使用した炭素材料銘柄毎で著しく異なり,最大16倍の違いが見られた.この違いは,試料の格子定数C_Oより求まる黒鉛化度によって説明することができ,高黒鉛化度の試料ほど水素溶解度が低くなることが分かった.また,中性子照射によっても水素溶解度が最大50倍に増加することから,水素は格子欠陥に捕捉されているものと考えられ,捕捉サイト濃度の評価を試みた.さらに,炭素-水素の強い結合が存在する系での脱捕獲,再捕獲,拡散,再結合を素過程とした動的な移動機構をモデル化して,実験結果との比較を行なった.今年度の成果については,これまでの実績と合わせて,第6回核融合炉材料国際会議(平成5年9月,イタリア),第2回核融合炉燃料-材料相互作用に関する日露ワークショップ(平成5年10月,ロシア)において発表を行なった.