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Yokoyama, Keisuke; Watanabe, Masashi; Onishi, Takashi; Yano, Yasuhide; Tokoro, Daishiro*; Sugata, Hiromasa*; Kato, Masato*
JAEA-Research 2025-002, 18 Pages, 2025/05
JAEA-Research-2025-002.pdf:1.73MB
It is advocated as a development target of fast reactors (FRs) to allow for the of use of mixed oxide (MOX) fuels containing minor actinide (MA) separated and recovered from spent fuels with the aim of reducing the volume and toxicity of high-level radioactive waste generated from nuclear reactors. In the development of MAMOX fuels, it is important behavior to understand the thermal properties such as thermal conductivity for fuel design and analysis of the irradiation. However, there are only a few reports on the thermal properties of MA-MOX fuels, and neither the effects of MA contents nor of oxygen non-stoichiometry in MOX fuels on their thermal conductivities have been fully understood. In this study, the thermal conductivities of MOX fuels with up to 15% Am content were measured at near-stoichiometric composition and the relationship between thermal conductivity and Am content was evaluated. Moreover, the thermal conductivities of Am-doped UO
fuels were also measured and evaluated by comparison with Am-MOX to evaluate the effect of Am content. The fuel samples used in this study were three types of MOX with a Pu content of 30% and different Am contents (5%, 10%, and 15%), and UO containing 15% Am. The thermal conductivities of specimens were calculated from the thermal diffusivities measured by the laser flash method, the density of the specimens and, the heat capacity at constant pressure. The oxygen partial pressure during the measurement was controlled at that of the targeted near-stoichiometric composition. The thermal conductivities of all specimens exhibited a decline with increasing temperature and Am content, with a particularly pronounced reduction observed below 1,173 K. The results of the classical phonon scattering model analysis of the measured thermal conductivities showed that the effect of lattice strain due to the Am addition was significant on the thermal resistivity change, and the effect was comparable for both MOX and UO.
Am
O at 1473, 1573, and 1673 KWatanabe, Masashi; Yokoyama, Keisuke; Vauchy, R.; Kato, Masato; Sugata, Hiromasa*; Seki, Takayuki*; Hino, Tetsushi*
Journal of Nuclear Materials, 599, p.155232_1 - 155232_5, 2024/10
Times Cited Count：2 Percentile：57.55(Materials Science, Multidisciplinary)Oxygen potential data of UAmO
were measured at 1473, 1573, and 1673 K by thermogravimetry. In U
An
O, where An stands for Pu or Am, and for a given value of y and Oxygen/Metal ratio, the oxygen potential of UAmO is higher than that of UPuO. The valence of cations in the hypostoichiometric region is similar to that of Nd-doped UO. At the stoichiometric composition, it is estimated to be Am
, U
, and U
(for charge compensation of Am). The experimental data were analyzed using a defect chemistry model, and a relationship connecting the oxygen-to-metal ratio, the temperature, and the equilibrium oxygen partial pressure was proposed.
Someya, Takayuki*; Chitose, Hiromasa*; Watanabe, Satoshi*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
In this study, CFD analysis has been conducted for the assessment of spent fuel integrity in large LOCA event and the maximum temperature of spent fuel assemblies has been evaluated. Then, it has been compared with the result of the simple assessment method. As a case study, additional CFD analysis has been conducted, where water level in SFP decreases to the Bottom of Active Fuel (BAF) due to boil-off. Since this scenario might be more severe than large LOCA scenario, the number of spent fuel assemblies, their decay heat and loading pattern to maintain spent fuel integrity are investigated.
-PuO-ZrO pseudo-ternary systemMorimoto, Kyoichi; Hirooka, Shun; Akashi, Masatoshi; Watanabe, Masashi; Sugata, Hiromasa*
Journal of Nuclear Science and Technology, 52(10), p.1247 - 1252, 2015/10
Times Cited Count：4 Percentile：29.40(Nuclear Science & Technology)As a part of decommissioning plan of the damaged reactors at Fukushima Daiichi Nuclear Power Plant, some strategies for removing of debris from the reactors are discussed. In these considerations, it is necessary to predict a melt progression during the severe accident based on theoretical evidences. Melting temperature is one of the most important thermal characteristics to analyse a melt progression during the severe accident. In this study, the melting temperatures of specimens of U, Pu and Zr mixed oxide prepared as simulated debris were measured by the thermal arrest technique. From the results of this measurement, the influences of Pu
and Zr contents on the melting temperature of the simulated debris were evaluated.
Tanaka, Susumu; Fukuda, Mitsuhiro; Nishimura, Koichi; Hosono, Masakazu; Watanabe, Hiromasa; Yamano, Naoki*
Journal of Nuclear Science and Technology, 37(Suppl.1), p.840 - 844, 2000/03
no abstracts in English
; Fukuda, Mitsuhiro; ; Watanabe, Hiromasa; Yamano, Naoki*
JAERI-Data/Code 97-019, 91 Pages, 1997/05
JAERI-Data-Code-97-019.pdf:2.58MB
no abstracts in English
Tanaka, Ryuichi; Yotsumoto, Keiichi; Watanabe, Hiromasa
Radioisotopes, 45(3), p.213 - 220, 1996/03
no abstracts in English
Watanabe, Hiromasa; Tanaka, Susumu; Nishimura, Koichi; Hosono, Masakazu
JAERI-Review 95-019, p.245 - 246, 1995/10
no abstracts in English
Watanabe, Hiromasa
Genshiryoku Kogyo, 40(2), p.12 - 15, 1994/00
no abstracts in English
Sekine, Toshiaki; Izumo, Mishiroku; Matsuoka, Hiromitsu; Kobayashi, Katsutoshi; Ishioka, Noriko; Osa, Akihiko; Koizumi, Mitsuo; Motoishi, Shoji; Hashimoto, Kazuyuki; ; et al.
Proc. of the 5th Int. Workshop on Targetry and Target Chemistry, 0, p.347 - 352, 1994/00
no abstracts in English
Tanaka, Susumu; Fukuda, Mitsuhiro; Nishimura, Koichi; Yokota, Wataru; Kamiya, Tomihiro; Watanabe, Hiromasa; Yamano, Naoki*; ;
Proc. of the 8th Int. Conf. on Radiation Shielding, 0, p.965 - 971, 1994/00
no abstracts in English
Nakamura, Yoshiteru; ; Nishimura, Koichi; Watanabe, Hiromasa; ; ;
Proc. of the 9th Symp. on Accelerator Science and Technology, p.434 - 436, 1993/00
no abstracts in English
Watanabe, Hiromasa; Tanaka, Susumu;
Hoken Butsuri, 26, p.395 - 404, 1991/00
no abstracts in English
; ; ; ; ; ; Iida, Hiromasa; Hoshiya, Taiji; ; ; et al.
Fusion Technology 1988, p.1806 - 1810, 1989/00
no abstracts in English
Oka, Kiyoshi; Nakamura, Tetsuya*; Ueda, Hirohisa*; Toriya, Tomoaki*; Tsumanuma, Koji*; Naganawa, Akihiro*; Watanabe, Shinsuke*; Ishiyama, Akihiko*; Yamashita, Hiromasa*; Chiba, Toshio*
no journal, ,
no abstracts in English
Hatakeyama, Shoichi*; Miura, Hiromasa*; Yao, Z.*; Tsutsui, Hiroaki*; Iio, Shunji*; Shibata, Yoshihide; Ono, Noriyasu*; Watanabe, Kiyomasa*; Akiyama, Tsuyoshi*; Nakamura, Kazuo*
no journal, ,
no abstracts in English
Chitose, Hiromasa*; Watanabe, Satoshi*; Sadamatsu, Hideaki*; Iwata, Yutaka*; Kaji, Yoshiyuki; Nemoto, Yoshiyuki
no journal, ,
We researched the recent enhancements of regulations and solutions related Spent Fuel Pool safety and reported the analytical results about problems and evaluation procedures related effectiveness evaluation of safety measures.
