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Iwamoto, Yosuke; Wakai, Eiichi; Nakagawa, Yuki*; Shibayama, Tamaki*
no journal, ,
In order to develop a new non-destructive inspection technique to accurately measure defects inside materials in radiation fields such as nuclear power plants, accelerators, and aerospace, we have measured the time evolution of electrical resistivity of copper, aluminum, and niobium metal wires and silicon plates by irradiating a pulsed laser beam (20 Hz) at Hokkaido University. From the obtained increase in electrical resistivity, the number of atomic vacancies and Frenkel pairs (FPs) of interstitial atoms and the dislocation density of the formed FPs were estimated, assuming that the atoms were ejected by the laser due to electronic excitation. From this experiment, it was estimated that for metals, the amount of defects formed inside the material increased with the increase in the irradiation dose. On the other hand, for silicon, the electrical resistivity was found to decrease due to the electronic transition to the valence band caused by laser irradiation.
Suzudo, Tomoaki; Ebihara, Kenichi; Tsuru, Tomohito; Mori, Hideki*
no journal, ,
BCC metals are used for various purposes as structural materials, but it is known that they become brittle in the low temperature range and brittleness is promoted by impurities such as hydrogen. It is critical to properly model and predict the phenomenon, but the mechanism is too complicated to understand. So far, we have performed molecular dynamics (MD) simulations of cleavage using the EAM empirical potential of iron and succeeded in reproducing a property of BCC metals, i.e., the cleavage propagating along {100}. However, plastic deformation at the crack tip, which is considered to be caused by the innaccuracy of the EAM potential, remained as a problem. In this study, we reexamined the fracture simulation using the newly developed potential that corrected the above problem.
Yano, Yasuhide; Toyama, Takeshi*; Tanno, Takashi; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Nakashima, Hideharu*; Onuma, Masato*; Kaito, Takeji
no journal, ,
no abstracts in English
Ohigashi, Jun*; Ueda, Mitsutoshi*; Kawamura, Kenichi*; Irisawa, Eriko; Komatsu, Atsushi; Kato, Chiaki
no journal, ,
In order to develop an oxygen sensor using high temperature oxide film of zirconium alloy as a solid electrolyte, an alloy of zirconium with nickel was investigated to suppress the growth rate of zirconia oxide film on the alloy. In this study, the high temperature oxidation behavior of zirconium-nickel alloy was evaluated. The alloy was oxidized at 873 K in a mixture of argon and 1% oxygen. The relationship between the difference in oxidation rate and in the properties of the oxide film were compared and discussed with the cross-sectional images of the oxide film. The oxidation rate was the same or slightly higher than that of pure zirconium. In addition, the oxide film became thicker and voids were formed in some areas, and some cracks were also observed. The results did not show any tendency to reduce the activity of zirconium in the alloy and to decrease the oxidation rate.
Toyota, Kodai; Wakai, Eiichi; Onizawa, Takashi; Shibayama, Tamaki*; Nakagawa, Yuki*
no journal, ,
no abstracts in English
Toyama, Takeshi*; Inoue, Koji*; Nagai, Yasuyoshi*; Kinomura, Atsushi*; Suzudo, Tomoaki; Hatano, Yuji*
no journal, ,
Tungsten (W) is promising as a fusion reactor plasma facing material. Retention and accumulation of hydrogen isotopes become a problem under irradiation, but it was found that the amount of hydrogen isotope accumulation is significantly reduced by the addition of rhenium (Re) and chromium (Cr). This is thought that Re and Cr suppress the formation of vacancy-type defects under irradiation, which are hydrogen trapping sites, but sufficient experimental findings to support this phenomenon have not been obtained yet. The purpose of this study is to apply the positron annihilation method and to investigate the effect of additive elements on the formation of irradiation defects in W alloys under electron beam and neutron irradiation.
Hiyoshi, Kensuke*; Egusa, Daisuke*; Yamaguchi, Masatake; Abe, Eiji*
no journal, ,
no abstracts in English
Yamazaki, Jin*; Onuma, Masato*; Tanno, Takashi; Otsuka, Satoshi; Toyama, Takeshi*; Mitsuhara, Masatoshi*; Nakashima, Hideharu*
no journal, ,
The stability of the nano-meter sized oxide dispersoids in ODS steel was evaluated by the small-angle X-ray scattering (SAXS) method in order to evaluate the properties of ODS steel under ultra-high temperatures assuming accident situations. It was found that the oxide is generally stable up to 1250 
C, but the oxide dispersion state changes even in a short time at 1300 C or higher. This theme was supported by JPMXD0219214482, a nuclear system research and development project of the Ministry of Education, Culture, Sports, Science and Technology.
Tanno, Takashi; Fujita, Koji; Yano, Yasuhide; Otsuka, Satoshi; Nakashima, Hideharu*; Mitsuhara, Masatoshi*; Onuma, Masato*; Toyama, Takeshi*; Kaito, Takeji
no journal, ,
By applying oxide dispersion-strengthened (ODS) steel, which has excellent high-temperature strength, to the fuel cladding tube, it is expected that the risk of fuel pin rupture up to the ultra-high temperature range including accidents will be reduced and the safety of the plant will be improved. Oxide dispersoids as a reinforced phase in ODS steels are considered to be more stable than carbonitrides and intermetallic compounds used as reinforced phases in other heat-resistant alloys. It is important to actually ensure that the metallographical structure and strength are maintained up to the ultra-high temperature. In this study, in order to evaluate the performance limit of ODS steel against ultra-high temperature, extreme heat treatments simulating accident situations were carried out, and then the macrostructure and strength were evaluated. Regarding the macrostructure, it was generally stable up to 1250 C and the crystal grains were fine, but it was coarsened at 1300 C or higher. On the other hand, it was found that the hardness decreased at 1250 C or higher. This theme was supported by JPMXD0219214482, a nuclear power system research and development project of the Ministry of Education, Culture, Sports, Science and Technology.
Tsuru, Tomohito; Lobzenko, I.; Han, S.*; Chen, Z.*; Kishida, Kyosuke; Inui, Haruyuki*
no journal, ,
In this study, the effect of the constituent elements on the mechanical properties of the ternary BCC medium entropy alloy (MEA) model was investigated by the first-principles calculation. For NbTiZr with different composition as BCC-MEA, we construct an atomic model with random solid solution and short-range ordered (SRO) structure obtained from Monte Carlo analysis. For each of the Random structure and SRO structure, various bulk properties related to the formation of the SRO, mean square atomic displacement (MSAD), elastic properties, stacking defects, twins, etc. are investigated. The formation energy and distribution of dislocation dipoles were evaluated by first-principles calculation of the dislocation structure, and the influence of Group 4 elements was examined.
Gong, W.; Zheng, R.*; Harjo, S.; Kawasaki, Takuro; Aizawa, Kazuya; Tsuji, Nobuhiro*
no journal, ,
Wakai, Eiichi; Iwamoto, Yosuke; Shibayama, Tamaki*; Sato, Koichi*; Toyota, Kodai; Onizawa, Takashi; Wakui, Takashi; Ishida, Taku*; Makimura, Shunsuke*; Nakagawa, Yuki*; et al.
no journal, ,
In the fields of accelerator target systems, nuclear power, aerospace, etc., radiation degradation of structural materials and equipment occurs, and therefore, the development of materials with high durability and excellent functions is expected. In order to create innovative materials that can be used in radiation fields, we are developing a new non-destructive inspection technique that can accurately measure the internal defects of various materials in radiation fields. As an innovative material, high-entropy alloys (HEA) are known for their high strength and ductility, and are expected to be used in various applications. In this talk, we will report on the construction of a measurement principle that enables multi-simultaneous measurements even in radiation fields, the status of HEA prototypes, and the status and progress of irradiation analysis of metals and other materials.
Wakai, Eiichi; Noto, Hiroyuki*; Kano, Sho*; Ishida, Taku*; Makimura, Shunsuke*; Shibayama, Tamaki*
no journal, ,
It is important to develop materials that can withstand high heat load and irradiation, and to evaluate their safety and lifetime for materials and devices used under high intensity beam in high energy accelerators, and for fusion reactor wall materials and divertors near high temperature plasmas. Tungsten-based materials are candidates as target materials in the second target station project of the Materials and Life Science Experimental Facility at the J-PARC center. In this study, we investigated the irradiation resistance of the materials fabricated by mixing tungsten with TiC particles of about 1.1 wt%, mechanical alloying, and high-temperature isostatic sintering. This material is an innovative nanoparticle-dispersed W material with a crystal grain size of about 1 to 2 
m and high strength. The irradiation resistance of this material and pure tungsten was examined by W ion irradiation at 773 K. These specimens were subjected to nanoindentation. These samples were analyzed by nanoindenter and transmission electron microscopy, and it was found for the first time that this nanoparticle-dispersed W material has a very high irradiation resistance compared to pure tungsten.