Experimental and modeling studies on the oxygen ingression behavior at the crevices of stainless steels in high-temperature water

Soma, Yasutaka; Komatsu, Atsushi; Kaji, Yoshiyuki; Yamamoto, Masahiro*; Igarashi, Takahiro

Corrosion Science, 251, p.112897_1 - 112897_15, 2025/07

https://doi.org/10.1016/j.corsci.2025.112897

Experimental and modeling studies of the oxygen ingression at the crevices of stainless steels were conducted in high-temperature water (288C). The limiting distance of oxygen ingression, , was defined as the point beyond which the primary surface oxide changed (hematite magnetite), regardless of crevice gap, oxygen concentration, and time. In situ measurements revealed increased electrical conductivity around the position indicating ion enrichment due to a differential oxygen concentration cell. increased with increasing crevice gap, oxygen concentration, and immersion time. Modeling study suggested that oxide layer growth reduced anodic dissolution and slowed oxygen consumption, allowing oxygen ingression with time.

Neutron diffraction study of the crystal and magnetic structures of antiferromagnetic manganese deuteride at high temperatures and high pressures

Machida, Akihiko*; Saito, Hiroyuki*; Aoki, Katsutoshi*; Komatsu, Kazuki*; Hattori, Takanori; Sano, Asami; Funakoshi, Kenichi*; Machida, Shinichi*; Sato, Toyoto*; Orimo, Shinichi*

Physical Review B, 111(22), p.224413_1 - 224413_6, 2025/06

https://doi.org/10.1103/yf8k-cy61

The crystal and magnetic structures of antiferromagnetic Mn deuterides formed by hydrogenating Mn metal at high temperature and high pressure, fcc -MnDx and hcp -MnDx, were investigated by in-situ neutron powder diffraction. Deuterium atoms partially occupied the octahedral interstitial positions of the fcc and hcp metal lattices. The site occupancies increased rapidly with decreasing temperature from 700 to 450 K and remained down to 300 K. Nel temperature of 543(10) K was determined for -MnD. For -MnD, saturation magnetic moment and Nel temperature were determined to be 0.82(1) and 347(3) K, respectively. The Nel temperatures determined for -MnD and -MnD are consistent with those predicted by the respective Slater-Pauling curves proposed in previous studies. The updated Nel temperatures provide insights into the development of more accurate Slater-Pauling curves based on electronic band structure calculations.

R&D status of HTGR heat utilization system and thermochemical H production IS process

Kubo, Shinji

Shokubai, 67(2), p.71 - 77, 2025/04

no abstracts in English

Forefront of development of next-generation innovative nuclear reactors (fast reactor and high-temperature gas-cooled reactor), 1; Latest trends of development of next-generation innovative nuclear reactors in Japan and foreign countries

Yamano, Hidemasa; Toyooka, Junichi; Sato, Hiroyuki; Sakaba, Nariaki

Nihon Genshiryoku Gakkai-Shi ATOMO, 66(12), p.607 - 611, 2024/12

https://doi.org/10.3327/jaesjb.66.12_607

This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries.

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.

Development of nuclear instruments to measure power distribution of HTGR, 1; Development of ex-core detector

Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Terao, Tsuyoshi*; Koike, Akifumi*

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

Japan Atomic Energy Agency, ANSeeN, and Shizuoka University has been conducted a joint-research to develop nuclear instrument for High Temperature Gas-cooled Reactor (HTGR) core power distribution for 3 years from 2021 supported by "Nuclear Energy System R&D Project" in MEXT. In the project, there are two R&Ds for "Development of ex-core detector" and "Development of in-core detector". The part of "Development of ex-core detector" is reported in this presentation. The "Development of ex-core detector" is innovative technology by virtue of long flight length neutron of graphite moderated HTGR core and Computed Tomography (CT) technologies. These technologies is expected to be applied to other reactors.

Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700C to 1000C

Imagawa, Yuya; Hashidate, Ryuta; Miyazawa, Takeshi; Onizawa, Takashi; Otsuka, Satoshi; Yano, Yasuhide; Tanno, Takashi; Kaito, Takeji; Onuma, Masato*; Mitsuhara, Masatoshi*; et al.

Journal of Nuclear Science and Technology, 61(6), p.762 - 777, 2024/06

https://doi.org/10.1080/00223131.2023.2269178

The Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650C to 850C. However, little data have been obtained above 850C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700C to 1000C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix's phase transformation, and a single equation can express a creep rupture strength from 700C to 1000C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.

Investigation of deposits on filter element of primary gas circulators in HTTR

Hasegawa, Toshinari; Nagasumi, Satoru; Nemoto, Takahiro; Nakajima, Kunihiro; Yokoyama, Keisuke; Fujiwara, Yusuke; Arakawa, Ryoki; Iigaki, Kazuhiko; Inoi, Hiroyuki; Kawamoto, Taiki

Proceedings of 2024 International Congress on Advanced in Nuclear Power Plants (ICAPP 2024) (Internet), 10 Pages, 2024/06

The filter element of the primary gas circulators (PGC) in High Temperature engineering Test Reactor (HTTR) and its deposits were investigated by Scanning Electron Microscope (SEM) observation and Energy Dispersive X-ray spectroscopy (EDX) analysis to find the cause of the increase of the filter differential pressure during the operation in 2021. SEM observation showed that the clumpy deposits and fibrous deposits smaller than the filtration pore size and the rod-shaped deposits larger than the pore size were present on the filter element. EDX analysis showed that the clumpy deposits and fibrous deposits could include silicone oil in the primary helium purification system (PHPS) gas circulators and that the rod-shaped deposits were thermal insulators inside of the co-axial double pipes in the primary cooling system. It is considered that silicone oil leaked from the PHPS gas circulators due to deterioration in the absorption performance of the activated charcoal filter. Next, it could be vaporized and reach PGC's filter element after passing through the reactor core. Since those deposits including silicone oil were present over the entire surface of the filter element, the filter differential pressure could be increased due to a reduction in the pore size and a rise in its flow resistance. The thermal insulator was unrelated to filter clogging because it was present mainly in the lower part of the filter element. Therefore, silicone oil could increase the filter differential pressure, and the graphite powder, which is the cause of the previous issue was unrelated.

High Temperature Gas-cooled Reactor (HTGR)

Noguchi, Hiroki; Sato, Hiroyuki; Nishihara, Tetsuo; Sakaba, Nariaki

Kagaku Kogaku, 88(5), p.211 - 214, 2024/05

High temperature gas-cooled reactor (HTGR), one of the next-generation innovative reactors, has an inherent safety and can generate very high-temperature heat which can be used for various heat application including hydrogen production. In Japan, Green Growth Strategy for Carbon Neutrality in 2050 and Basic Policy for the Realization of GX state the promotion of technology development necessary for mass and low-cost carbon-free hydrogen production and development and construction of next-generation innovative reactors including the HTGR for the decarbonization of industrial sectors. Based on these policies, JAEA has been conducted the world's first hydrogen production test using nuclear heat from an HTGR, in addition to verifying the excellent safety features of HTGR, and has also started to study the construction of an HTGR demonstration reactor in cooperation with the industrial community. This paper shows the current status of R&D of HTGR in Japan.

Technology information on High Temperature Gas-cooled Reactor (HTGR)

HTGR Design Group, HTGR Project Management Office

JAEA-Technology 2023-019, 39 Pages, 2024/01

JAEA-Technology-2023-019.pdf:1.34MB

In order to realize the development of the demonstration reactor of High Temperature Gas-cooled Reactor (HTGR) with a target of starting operation in the 2030s, as indicated in the "Basic Policy for GX Realization" (Cabinet Decision on February 10, 2023) and the Working Group on Innovative Reactors of METI, Japan Atomic Energy Agency (JAEA) has been working on the development of a standard for the development of a HTGR under the Atomic Energy Society of Japan and the Japan Society of Mechanical Engineers. In addition, JAEA has been commissioned by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry (METI) to conduct the "Demonstration Project for Mass Hydrogen Production Technology Using Ultra-High Temperatures" and has been promoting a hydrogen production project using the HTTR (High Temperature Engineering Test Reactor). Furthermore, in collaboration with the National Nuclear Laboratory (NNL) of the United Kingdom and the National Centre for Nuclear Research (NCBJ) of Poland, JAEA are aiming to strengthen the international competitiveness of HTGR technology by further upgrading the HTGR technology developed in Japan through the construction and operation of the HTTR. In response to the growing interest in HTGR development in Japan and abroad, we have developed FAQs on HTGR related technologies in order to provide accurate technical information on HTGRs.