Complementary layer thickness effects of Q235 and SUS304 layers of multilayered steels for improving of tensile strength and plasticity simultaneously

Lin, Z. M.*; Liu, B. X.*; Ming, K. S.*; Xu, P. G.; Yin, F. X.*; Zheng, S. J.*

Scripta Materialia, 263, p.116692_1 - 116692_7, 2025/07

https://doi.org/10.1016/j.scriptamat.2025.116692

Chapter 9, Advanced materials; Oxide-dispersion strengthened steels

Otsuka, Satoshi; Tanno, Takashi; Yano, Yasuhide; Kaito, Takeji

Materials and Processes for Nuclear Energy Today and in the Future, p.279 - 297, 2024/10

https://doi.org/10.1002/9781394325870.ch9

The oxide dispersion strengthening is an effective technique for improving the mechanical strength of the steel. The dispersed oxides prevent the gliding motion of dislocations, thus remarkably enhancing the resistance to high-temperature deformation and rupture of steels. Extensive efforts have been made to develop ODS steels in the fields of nuclear and fusion engineering. Research has been done to improve their performance and meet the requirements such as irradiation resistance, high-temperature strength, and corrosion resistance. Based on recent research, the high-density dispersion of nanosized oxides could improve the irradiation resistance of the steels in addition to high-temperature strength because the interface between oxide and matrix could act as sink sites for point defects. This section overviews the ODS steel development for nuclear application.

Influence of interstitial carbon on bulk texture evolution of carbide-free high-entropy alloys during cold rolling using neutron diffraction

Fang, W.*; Liu, C.*; Zhang, J.*; Xu, P. G.; Peng, T.*; Liu, B.*; Morooka, Satoshi; Yin, F.*

Scripta Materialia, 249, p.116046_1 - 116046_6, 2024/08

https://doi.org/10.1016/j.scriptamat.2024.116046

Microstructures and hardness of BCC phase iron-based high entropy alloy Fe-Mn-Cr-V-Al-C

Wakai, Eiichi; Noto, Hiroyuki*; Shibayama, Tamaki*; Furuya, Kazuyuki*; Ando, Masami*; Kamada, Takaharu*; Ishida, Taku*; Makimura, Shunsuke*

Materials Characterization, 211, p.113881_1 - 113881_10, 2024/05

https://doi.org/10.1016/j.matchar.2024.113881

The microstructures and mechanical properties of bcc iron-based high entropy alloy (HEA) Fe-20Mn-15Cr-10V-10Al-2.5C (in at%) without Co and Ni elements have been investigated for applications in fields such as accelerator-target system, nuclear reactors and magnetic motors in aircraft and automobiles. This alloy was normalized at 1150C for 2 hr and then water quenched, and it was heated at 800C for 10 min and then water quenched. The alloy had a bcc-phase and vanadium carbides with 2-3 m arranging along grain boundaries, and the Vickers hardness was 520 Hv, harder than pure tungsten. Magnetic domain structure was observed in phase differential contrast method in scanning transmission electron microscope, and the micro-size magnetic domains in grain and sub micro size ones were formed near surface, and it is attractive to the magnetic motor field application. Element distribution in nano scale (20 nm) was observed in matrix, and the presence of crystal lattice disorder in the atomic level region was seen. Very high performance for radiation resistance was confirmed with no irradiation hardening at 300 and 500 C to 1 dpa. It can be speculated that this is due to irradiation-induced nanoscale concentration changes and strain relaxation in the HEA. These properties are very attractive in application of several fields.

Recent improvement and evaluation of radiation resistance and magnetic properties of high entropy alloys and their applications

Wakai, Eiichi; Noto, Hiroyuki*; Shibayama, Tamaki*; Furuya, Kazuyuki*; Wakui, Takashi; Ando, Masami*; Makimura, Shunsuke*; Ishida, Taku*

Science Talks (Internet), 8, p.100278_1 - 100278_4, 2023/12

https://doi.org/10.1016/j.sctalk.2023.100278

High entropy alloys tend to combine high strength with good ductility due to their inherent properties. This material is considered as a promising new material not only for higher-performance future general industrial applications, but also for increasing the durability and range of application of radiation-affected equipment in nuclear and radiation environments, and has been rapidly gaining attention in recent years. In this study, two types of high-entropy alloys (Fe-Mn-V-Cr-Al-C and Fe-Si-W-Cr-V) composed of low-radioactive elements (without Ni and Co) were prepared and their basic properties were evaluated for application as new functional materials to be used under radiation in high-energy accelerator target system components, nuclear reactors, fusion reactors, etc. and their basic properties were evaluated. The two materials under development in this study have unique properties in the following respects. The former is expected to be developed as a basic research for high-power motor materials as a new structural material and magnetic properties sharing the features of high strength and low radiation. On the other hand, the latter is expected to be applied as a new functional material in new engineering fields by mixing tungsten, which has the highest melting point among metallic elements, with vanadium, which has a considerably higher melting point, to raise the melting point of the alloy and to design an alloy with high strength.

Study on the uptake of Sr and Cs into wild animals after the Fukushima Daiichi Nuclear Power Plant accident using hard tissue as an indicator

Koarai, Kazuma

Hosha Kagaku, (47), p.24 - 27, 2023/03

Ca-based tissues, such as teeth and bones, are called hard tissues. Sr is incorporated into the tooth during calcification. Once incorporated, it remains until the tooth falls out. Therefore, Sr in teeth recorded Sr contamination by the 1F accident in the environment during the formation period. The article presents current research findings.

High temperature mechanical properties and microstructure in 9Cr or 12Cr oxide dispersion strengthened steels

Mitsuhara, Masatoshi*; Kurino, Koichi*; Yano, Yasuhide; Otsuka, Satoshi; Toyama, Takeshi*; Onuma, Masato*; Nakashima, Hideharu*

Tetsu To Hagane, 109(3), p.189 - 200, 2023/03

https://doi.org/10.2355/tetsutohagane.TETSU-2022-080

Oxide Dispersion Strengthened (ODS) ferritic steel, a candidate material for fast reactor fuel cladding, has low thermal expansion, good thermal conductivity, and excellent resistance to irradiation damage and high temperature strength. The origin of the excellent high-temperature strength lies in the dispersion of fine oxides. In this study, creep tests at 700 or 750C, which are close to the operating temperatures of fast reactors, and high-temperature tensile tests at 900 to 1350 C, which simulate accident conditions, were conducted on 9Cr ODS ferritic steels, M11 and MP23, and 12Cr ODS ferritic steel, F14, to confirm the growth behavior of oxides. In the M11 and F14 creep test samples, there was little oxide growth or decrease in number density from the initial state, indicating that dispersion strengthening by oxides was effective during deformation. After creep deformation of F14, the development of dislocation substructures such as dislocation walls and subgrain boundaries was hardly observed, and mobile dislocations were homogeneously distributed in the grains. The dislocation density increased with increasing stress during the creep test. In the high-temperature ring tensile tests of MP23 and F14, the strength of both steels decreased at higher temperatures. In MP23, elongation decreased with increasing test temperature from 900 to 1100 C, but increased at 1200 C, decreased drastically at 1250 C, and increased again at 1300 C. In F14, elongation decreased with increasing temperature. It was inferred that the formation of the -ferrite phase was responsible for this complex change in mechanical properties of MP23 from 1200 to 1300 C.

While acknowledging each other's strengths and weaknesses, let's support each other and create the future

Oba, Kyoko

Nihon Genshiryoku Gakkai-Shi ATOMO, 64(11), P. 646, 2022/11

https://doi.org/10.3327/jaesjb.64.11_646

This report was written the final part of the relay series commemorating the 20th anniversary of the Ethical Code of the Atomic Energy Society of Japan. As chairperson, I expresses my thoughts on the future of ethics.

Radiation chemistry provides nanoscopic insights into the role of intermediate phases in CeO mesocrystal formation

Li, Z.*; Piankova, D.*; Yang, Y.*; Kumagai, Yuta; Zschiesche, H.*; Jonsson, M.*; Tarakina, N. V.*; Soroka, I. L.*

Angewandte Chemie; International Edition, 61(6), p.e202112204_1 - e202112204_9, 2022/02

https://doi.org/10.1002/anie.202112204

The role of intermediate phases in CeO mesocrystal formation from aqueous Ce(III) solutions subjected to -radiation was studied. Radiolytically formed hydroxyl radicals convert soluble Ce(III) into less soluble Ce(IV). Transmission electron microscopy and X-ray diffraction studies of samples from different stages of the process allowed the identification of several stages in CeO mesocrystal evolution following the oxidation to Ce(IV): (1) formation of hydrated Ce(IV)-hydroxides, serving as intermediates in the liquid-to-solid phase transformation; (2) CeO primary particle growth inside the intermediate phase; (3) alignment of the primary particles into "pre-mesocrystals" and subsequently to mesocrystals, guided by confinement of the amorphous intermediate phase and accompanied by the formation of mineral bridges. Further alignment of the obtained mesocrystals into supracrystals occurs upon slow drying, making it possible to form complex hierarchical architectures.

Development of fuel performance analysis code, BISON for MOX, named Okami; Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

Ozawa, Takayuki; Hirooka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*

Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09

AA2020-0710.pdf:1.77MB

https://doi.org/10.1016/j.jnucmat.2021.153038

To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.