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Saito, Toshiki*; Yu, H.*; Gao, Z.*; Inoue, Koji*; Kondo, Sosuke*; Kasada, Ryuta*; Nagai, Yasuyoshi*; Oba, Yojiro*; Hiroi, Kosuke
Journal of Alloys and Compounds, 983, p.173852_1 - 173852_12, 2024/05
Haoran, W.*; Yu, H.*; Liu, J.*; Kondo, Sosuke*; Okubo, Nariaki; Kasada, Ryuta*
Corrosion Science, 209, p.110818_1 - 110818_12, 2022/12
Times Cited Count：17 Percentile：77.23(Materials Science, Multidisciplinary)The corrosion behavior of newly developed Al
O
forming high Mn oxide dispersion strengthened (ODS) austenitic steels was examined in oxygen-saturated lead-bismuth eutectic at 450
C for 430 h. Compared with non-ODS steels, the ODS steels possessed superior resistance to corrosion and spallation. The high density grain boundaries in the ODS steels acted as channels for the rapid outward diffusion of metallic elements, forming an internal continuous CrO scale at the original surface. Accelerated Al diffusion, along with oxidation prevention by the external (Fe, Mn) oxide scale and the internal CrO scale, jointly resulted in the formation of a continuous Al-rich oxide scale in ODS-7Al steel, contributing to its superior corrosion resistance.
CWang, H.*; Yu, H.*; Kondo, Sosuke*; Okubo, Nariaki; Kasada, Ryuta*
Corrosion Science, 175, p.108864_1 - 108864_12, 2020/10
Times Cited Count：47 Percentile：91.13(Materials Science, Multidisciplinary)Corrosion tests were performed on newly developed alumina-forming austenitic (AFA) steels in stagnant lead bismuth eutectic (LBE) with saturated and low oxygen concentrations at 450C for 430 h. The steels exhibited enhanced corrosion resistance to the LBE environments with the increasing of Al content. A continuous and protective Al-rich oxide scale formed on the steel specimens that were exposed to LBE with a low oxygen concentration, whereas a non-protective and stratified oxide scale formed in the oxygen saturated LBE.
' precipitates in nickel based oxide dispersion-strengthened superalloys under high-temperature and heavy irradiation conditionsKonno, Azusa; Ono, Naoko*; Ukai, Shigeharu*; Kondo, Sosuke*; Hashitomi, Okinobu*; Kimura, Akihiko*
Materials Transactions, 60(11), p.2260 - 2266, 2019/11
Times Cited Count：1 Percentile：4.46(Materials Science, Multidisciplinary)A stability of cuboidal ' phase under heavy irradiation was studied for newly developed Ni-based Oxide Dispersion Strengthened (ODS) superalloy as a candidate for the core structural materials in VHTR or GFR. The ion irradiation was applied at 873K, 1073K, 1273K and the dose was 100 dpa. The ' phase remained the cuboidal shape at 873K but got out of the shape at 1073K after irradiation. Those growths can be explained by the Nelson-Hudson-Mazey (NHM) model. For the result of 1273K irradiation, however, huge ' phase appeared in the whole irradiated area at the post irradiation-observation. This behavior is interpreted in terms of disordering of the ordered ' phase due to cascade collision, and thus increasing Gibbs free energy of the disordered phase induces a change of the element distribution inside the irradiated area. The ordered ' phase was reproduced from the disordered state at the cooling after ion irradiation.
Kasada, Ryuta*; Wang, H.*; Liu, J.*; Yu, H.*; Kondo, Sosuke*; Okuno, Yasuki*; Okubo, Nariaki; Tokunaga, Toko*; Ono, Naoko*
no journal, ,
With the aim of applying to materials in accelerator driven systems (ADS) core and advanced blankets of fusion systems, a Fe-Mn-Cr-Al-C steel, low-activation austenitic steel, and its oxide dispersion-strengthened (ODS) alloy have been developing. By reviewing the composition of Fe-Cr-Mn-based low-activation austenitic steel that was once examined, the ODS alloy has become a material design that is conscious of twinning induced plasticity (TWIP) steel used as an automobile steel plate, and it aims to improve high temperature strength characteristics and irradiation resistance by ODS. This presentation shows the initial result of the influence of ODS on the strength characteristics and lead bismuth corrosion on the alloy design policy and materials prototyped at the laboratory level. In addition, some issues for application to advanced fusion system blankets are also discussed.
Saito, Toshiki*; Yu, H.*; Inoue, Koji*; Zimo, G.*; Kondo, Sosuke*; Kasada, Ryuta*; Nagai, Yasuyoshi*; Oba, Yojiro; Hiroi, Kosuke
no journal, ,
no abstracts in English
Saito, Toshiki*; Yu, H.*; Inoue, Koji*; Zimo, G.*; Kondo, Sosuke*; Kasada, Ryuta*; Nagai, Yasuyoshi*; Oba, Yojiro; Hiroi, Kosuke
no journal, ,
no abstracts in English
Soma, Yasutaka; Yamashita, Shinichiro; Hasegawa, Akira*; Kondo, Sosuke*
no journal, ,
The In-pile tests for Cr-Zry, which is one of the candidate materials for accident tolerant fuels, is currently dependent on overseas reactors. To improve this situation, JAEA and Tohoku University are developing an alternative irradiation corrosion evaluation technique, the "Proton Irradiation Corrosion Test Facility," which can be more flexibly conducted domestically. This facility, located at the Research Center for Accelerator and Radioisotope Science (RARiS) of Tohoku University, aims to evaluate the behavior of Cr-Zry under a combined environment of proton beam irradiation, corrosion in high-temperature water, and water radiolysis, simulating reactor conditions. The development of the facility began in 2022, and irradiation testing is planned to be carried out within this fiscal year. This presentation will report on the progress of the facility's development.
Saito, Toshiki*; Yu, H.*; Inoue, Koji*; Zimo, G.*; Kondo, Sosuke*; Nagai, Yasuyoshi*; Oba, Yojiro; Hiroi, Kosuke; Kasada, Ryuta*
no journal, ,
Soma, Yasutaka; Yamashita, Shinichiro; Hasegawa, Akira*; Kondo, Sosuke*
no journal, ,
Cr-coated cladding (Cr-Zry) is subjected to degradation due to corrosion and other factors in the operating environment, but published data is insufficient. Currently, in-pile tests for the performance evaluation of Cr-Zry depend on overseas reactors, which requires large amount of cost and time despite small number of output data. Therefore, an alternative evaluation technique under irradiation that can be performed more flexibly in Japan is strongly needed. In response to this need, JAEA has been developing a device that can evaluate the behavior of Cr-Zry under the effect of simultaneous irradiation and corrosion. In this presentation, we report on the development status of the apparatus equipped with a system for the in-situ measurement of corrosion behavior by electrochemical measurement, which is a feature not found in previous studies.
Ozawa, Kazumi; Koyanagi, Takaaki*; Taguchi, Tomitsugu; Nozawa, Takashi; Tanigawa, Hiroyasu; Kondo, Sosuke*; Hinoki, Tatsuya*
no journal, ,
A SiC/SiC composite is a promising candidate functional/structural material for fusion DEMO reactor. To examine the effects of transmuted H mainly on microstructural cavity formation, the monolithic NITE-SiC with 6wt% YO-AlO sintering additive system were ion-irradiated to 10 dpa at 1000C nominally with 130 appmHe/dpa, and 40 or 400 appmH/dpa, respectively. In SiC grains, tiny cavities with 2 nm formed densely. However, it is revealed that H could not remarkably impact on cavity formation (size, density) in the condition studied. The influence of cavities formed along SiC-YAG grain boundary (GB) on cavity swelling seems to be also small, possibly attributed to its quite low density, even though the cavity size formed is three to five times larger than the size in CVI-SiC matrix in the same irradiation condition. Additional investigation of the microstructural evolution in YAG grain and SiC-YAG GB is a future plan.
Ozawa, Kazumi; Koyanagi, Takaaki*; Taguchi, Tomitsugu; Nozawa, Takashi; Tanigawa, Hiroyasu; Kondo, Sosuke*; Hinoki, Tatsuya*
no journal, ,
no abstracts in English
Ozawa, Kazumi; Kondo, Sosuke*; Nozawa, Takashi; Tanigawa, Hiroyasu; Hinoki, Tatsuya*
no journal, ,
A SiC/SiC composite is a promising material for fusion DEMO reactor. In this presentation, dimensional and microstructural stabilities of advanced SiC fibers after high-dose ion irradiation at the temperature, where the composite is to be used as a flow channel insert (FCI), were evaluated by step height measurement and FE-TEM, respectively. As a preliminary result using the step height measurement, matrix and fiber was relatively flat and had no significant gap in a Tyranno-SA3 composite after ion irradiation at 600C to 100 dpa, while in a Hi-Nicalon Type-S composite a hollow part was observed in the center of the fiber. Differences of two types of advance SiC fibers on microstructure were discussed, compared with previous neutron/ion- irradiated data, considering the issues associated with ion irradiation and effects of pyloritic carbon interphase.
Ozawa, Kazumi; Koyanagi, Takaaki*; Nozawa, Takashi; Kato, Yutai*; Kondo, Sosuke*; Tanigawa, Hiroyasu; Snead, L. L.*
no journal, ,
A silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite is a promising candidate material for an advanced fusion DEMO blanket. High-dose irradiation experiments were performed with our special focuses on understanding; (1) integrity of the Hi-Nicalon Type-S (HNLS) composites, (2) functionality of thin pyrocarbon (PyC) /SiC multilayer, and (3) clarifying the mechanism underlying degradation, as feedback to R&D on SiC/SiC composites. The materials used in this study were plain-weave HNLS composites produced via the chemical vapor infiltration process. Neutron irradiation was conducted in the HFIR at ORNL. The peak neutron fluence was ~1.0
10
n/m
(E 
0.1 MeV, equivalent to ~100 dpa) at nominal irradiation temperatures of 300, 500, and 800C. Results of post irradiation experiments including 1/4-four-point flexural tests, SEM, and TEM observation were reported.
Ozawa, Kazumi; Koyanagi, Takaaki*; Nozawa, Takashi; Kato, Yutai*; Kondo, Sosuke*; Tanigawa, Hiroyasu; Snead, L. L.*
no journal, ,
A silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite is a promising candidate material for an advanced fusion DEMO blanket because of the excellent thermo-mechanical / -chemical properties and irradiation tolerance of SiC itself. The irradiation response of highly-crystalline and near-stoichiometric SiC fiber has been assumed to be the same as that of high purity monolithic 
-SiC. Unfortunately, based on recent data this assumption appears not to be correct. This study mainly aims to investigate mechanical and microstructural changes of the SiC/SiC composite after neutron irradiation to higher dose. In the post neutron irradiation experiments to 100 dpa, ~50% degradation of proportional limit stress and ultimate flexural strength was observed for the composite irradiated at 629C, while the one irradiated at 319C exhibited brittle behavior. In the ion-experiment results, shrinkage of the HNLS fiber was observed for specimens irradiated at 300 and 600C to 100 dpa. The FE-TEM results, EELS analysis and previous our works suggest that the shrinkage would be responsible for the transport of excess carbon atoms from intergranular phase into SiC grains.
Soma, Yasutaka; Yamashita, Shinichiro; Hasegawa, Akira*; Kondo, Sosuke*; Ito, Masatoshi*; Iwamoto, Chihiro*
no journal, ,
The development of Accident Tolerant Fuel (ATF, also referred to as Advanced Technology Fuel) requires irradiation testing in research reactors. However, such reactor irradiations depend heavily on overseas facilities and involve issues such as high cost and limited accessibility. Therefore, the development of alternative accelerator-based irradiation technologies is needed. In this study, we developed an accelerator-based irradiation test system for ATF material research. The system was jointly developed with the Institute for Materials Research (IMR), Tohoku University, and the Research Center for Accelerator and Radioisotope Science (RARiS), Tohoku University, and the resulting device-the Thin-Window Simultaneous Irradiation-Corrosion (T-SIC) system-was installed at RARiS. The T-SIC enables simultaneous proton irradiation and high-temperature water corrosion, offering the potential to reproduce radiation-induced oxidation and dissolution phenomena that cannot be realized in conventional sequential tests. The system is designed for Cr-coated Zircaloy specimens and will also allow in-situ electrochemical measurements under irradiation. In March 2025, irradiation tests using 20 MeV protons were conducted. Beam transport was successfully achieved, and irradiation of the specimen was confirmed, thereby demonstrating the feasibility of the system. Moving forward, improvements in system performance will be pursued, together with short-term irradiations to evaluate specimen activation, and verification of in-situ measurement capabilities.
Serizawa, Hiroyuki; Kondo, Sosuke*; Hinoki, Tatsuya*
no journal, ,
Pieces of CeO (100) thin film were heat treated at 1273 K for 2h. The films were irradiated with 130-keV He
ions using 400-keV ion implanter of TIARA. The ion doped film was heat treated at 1773 K for 2 h in air. After the heat treatment, the sample for STEM analysis was prepared for STEM analysis by FIB. It was found that many blisters were formed on the surface of the thin film. The blisters are considered to be formed by gas babble accompanied by the precipitation of He beneath the surface. The lid of the blister is blown away since the sample is ceramics. Many gas babbles are formed in the thin film. The size of the gas bubble falls within the range from 30 to 100 nm in diameter. The shape of the gas bubble is truncated octahedron but clearly different from that of void, which mean that the existence of He in the gas bubble effect on the shape of the gas bubble.
Konno, Azusa; Ono, Naoko*; Ukai, Shigeharu*; Kondo, Sosuke*; Hashitomi, Okinobu*; Kimura, Akihiko*
no journal, ,
A newly developed Ni-based Oxide Dispersion Strengthened (ODS) superalloy as a candidate for the core structural materials in VHTR or GFR is studied to evaluate the stability of cuboidal-gamma prime phase under heavy irradiation in the presence of oxide particles. The ion irradiation was applied at 873K, 1073K, 1273K and the dose was 100 dpa. The gamma prime phase kept the cuboidal shape at 873K and got out of the shape at 1073K after irradiation, where these growths can be explained by the NHM model. For the result of 1273K irradiation, where huge gamma prime phase appeared in the whole irradiated area, changing Gibbs free energy by irradiation could induce the order-disorder transformation.
precipitates in nickel based oxide dispersion-strengthened superalloys under severe conditionsKonno, Azusa; Ono, Naoko*; Ukai, Shigeharu; Kondo, Sosuke*; Hashitomi, Okinobu*; Kimura, Akihiko*
no journal, ,
It is essential to develop the structural materials in Very High-temperature Reactors (VHTR) or Gas-cooled Fast Reactors (GFR) which reactor core environment is severe as the temperature is over 1273 K and irradiation level is up to 100 dpa. In the past, Ni-based alloys which are superior to Fe-based alloys as high-temperature strength were studied because of its precipitates for reactor core materials. However, dissolves and reprecipitates at grain boundaries by irradiation at about 873 K. In order to show the suppression of this embrittlement, we newly developed a precipitation type Ni-based Oxide Dispersion-Strengthened (ODS) superalloy in which nano-sizeed oxide particles are finely dispersed. In this research, the stability of cuboidal precipitates under heavy irradiation was studied for newly developed Ni-based ODS superalloys, to explore the suitability of these as core materials. The specimen composition was equivalent to the MS4 produced by modifying the commercial superalloy CMSX10 to which the oxide particles are added.
