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Luu, V. N.; Taniguchi, Yoshinori; Udagawa, Yutaka; Katsuyama, Jinya
Nuclear Engineering and Design, 442, p.114222_1 - 114222_15, 2025/10
neutron diffraction studyYamashita, Takayuki*; Koga, Norimitsu*; Mao, W.*; Gong, W.; Kawasaki, Takuro; Harjo, S.; Fujii, Hidetoshi*; Umezawa, Osamu*
Materials Science and Engineering A, 941, p.148602_1 - 148602_11, 2025/09
Mao, W.*; Gong, W.; Kawasaki, Takuro; Gao, S.*; Ito, Tatsuya; Yamashita, Takayuki*; Harjo, S.; Zhao, L.*; Wang, Q.*
Scripta Materialia, 264, p.116726_1 - 116726_6, 2025/07
Times Cited Count：0
C alloy measured in the electrostatic levitation furnace onboard the international space stationIshikawa, Takehiko*; Oda, Hirohisa*; Koyama, Chihiro*; Shimonishi, Rina*; Ikeuchi, Rumiko*; Paradis, P.-F.*; Okada, Jumpei*; Fukuyama, Hiroyuki*; Yamano, Hidemasa
International Journal of Microgravity Science and Application, 42(2), p.420202_1 - 420202_10, 2025/04
Sakurai, Junya*; Torigata, Keisuke*; Matsunaga, Manabu*; Takanashi, Naoto*; Hibino, Shinya*; Kizu, Kenichi*; Morita, Akira*; Inomoto, Masahiro*; Shimohata, Nobuaki*; Toyota, Kodai; et al.
Tetsu To Hagane, 111(5), p.246 - 262, 2025/04
Aihara, Jun; Ueta, Shohei; Honda, Masaki*; Kasahara, Seiji; Okamoto, Koji*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for the purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO
(PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. In research project of Pu-burner HTGR carried out from fiscal year of 2014 to fiscal year of 2017, simulated CFPs were fabricated using Ce to simulate Pu. Moreover, simulated fuel compacts were fabricated using fabricated simulated CFPs. In this report, results of microstructural observation of CeO-YSZ and ZrC layer at each fabrication step are reported.
Miura, Taito*; Miyamoto, Shintoro*; Maruyama, Ippei*; Aili, A.*; Sato, Takumi; Nagae, Yuji; Igarashi, Go*
Case Studies in Construction Materials, 21, p.e03571_1 - e03571_14, 2024/12
Times Cited Count：0 Percentile：0.00(Construction & Building Technology)Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
JAEA-Technology-2024-009.pdf:8.03MB
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900
C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
Mao, W.*; Gao, S.*; Gong, W.; Kawasaki, Takuro; Ito, Tatsuya; Harjo, S.; Tsuji, Nobuhiro*
Acta Materialia, 278, p.120233_1 - 120233_13, 2024/10
Times Cited Count：12 Percentile：87.44(Materials Science, Multidisciplinary)Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
International Journal of Pressure Vessels and Piping, 211, p.105298_1 - 105298_6, 2024/10
Lan, Z.*; Arikawa, Yasunobu*; Mirfayzi, S. R.*; Morace, A.*; Hayakawa, Takehito*; Sato, Hirotaka*; Kamiyama, Takashi*; Wei, T.*; Tatsumi, Yuta*; Koizumi, Mitsuo; et al.
Nature Communications (Internet), 15, p.5365_1 - 5365_7, 2024/07
Times Cited Count：3 Percentile：79.20(Multidisciplinary Sciences)Shimazaki, Yosuke; Jidaisho, Tatsuya; Ishii, Toshiaki; Inoi, Hiroyuki; Iigaki, Kazuhiko
JAEA-Technology 2024-005, 23 Pages, 2024/06
JAEA-Technology-2024-005.pdf:5.53MB
HTTR has newly assumed Beyond Design Basis Accident (BDBA) as part of conformity assessment with the new regulatory standards and has established measures to prevent the spread of BDBA. Among these measures, to prevent the spread of BDBA caused by cooling water leaks from spent fuel storage pool, the Oarai Research Institute's fire engine was selected as an equipment to prevent the spread of BDBA, and required performances such as pumping water performance were determined. After all required performances were confirmed by inspections, the fire engine passed the operator's pre-use inspection and contributed to the restart of the HTTR operations.
Amaya, Masaki
High Temperature Corrosion of Materials, 101(3), p.455 - 469, 2024/06
Times Cited Count：0 Percentile：0.00(Metallurgy & Metallurgical Engineering)Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.
Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05
Times Cited Count：1 Percentile：57.00(Materials Science, Multidisciplinary)Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Zhang, Y.*; Marusawa, Kenji*; Kudo, Kohei*; Morooka, Satoshi; Harjo, S.; Miyamoto, Goro*; Furuhara, Tadashi*
ISIJ International, 64(2), p.245 - 256, 2024/01
Times Cited Count：2 Percentile：63.37(Metallurgy & Metallurgical Engineering)Hirota, Noriaki; Nakano, Hiroko; Fujita, Yoshitaka; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Demura, Masahiko*; Kobayashi, Yoshinao*
The IV International Scientific Forum "Nuclear Science and Technologies"; AIP Conference Proceedings 3020, p.030007_1 - 030007_6, 2024/01
Dynamic strain aging (DSA) and intergranular stress corrosion cracking (intragranular SCC) occur in high temperature pressurized water simulating a boiling water reactor environment due to changes in dissolved oxygen (DO) content, respectively. In order to clearly understand the difference between these phenomena, the mechanism of their occurrence was summarized. As a result, it was found that DSA due to intragranular cracking occurred in SUS304 stainless steel at low DO 
1 ppb, while DSA was suppressed at DO 100 to 8500 ppb due to the formation of oxide films on the surface. On the other hand, when DO was increased to 20000 ppb, the film was peeled from the matrix, O element diffused to the grain boundary of the matrix, resulting in intergranular SCC. These results are indicated that the optimum DO concentration must be adjusted to suppress crack initiation due to DSA and intergranular SCC.
CMohamad, A. B.; Nemoto, Yoshiyuki; Furumoto, Kenichiro*; Okada, Yuji*; Sato, Daiki*
Corrosion Science, 224, p.111540_1 - 111540_15, 2023/11
Times Cited Count：5 Percentile：55.06(Materials Science, Multidisciplinary)Kamide, Hideki; Kawasaki, Nobuchika; Hayafune, Hiroki; Kubo, Shigenobu; Chikazawa, Yoshitaka; Maeda, Seiichiro; Sagayama, Yutaka; Nishihara, Tetsuo; Sumita, Junya; Shibata, Taiju; et al.
Jisedai Genshiro Ga Hiraku Atarashii Shijo; NSA/Commentaries, No.28, p.14 - 36, 2023/10
Developments of next generation nuclear reactors, e.g., Fast Reactor, and High Temperature Gas cooled Reactor, are in progress. They can contribute to markets of electricity and industrial heat utilization in the world including Japan. Here, current status of reactor developments in Japan and also situation in the world are summarized, especially for activities of Generation IV International Forum (GIF), developments of Fast Reactor and High Temperature Gas cooled Reactor in Japan, and SMR movements in the world.
Wang, Q.*; Ma, N.*; Huang, W.*; Shi, J.*; Luo, X.-T.*; Tomitaka, Sora*; Morooka, Satoshi; Watanabe, Makoto*
Materials Research Letters (Internet), 11(9), p.742 - 748, 2023/09
Times Cited Count：2 Percentile：25.24(Materials Science, Multidisciplinary)