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Okamoto, Naritoshi; Komeno, Akira; Seya, Atsumasa; Inaba, Hideki*; Terakado, Shinichi*; Higuchi, Masashi*
JAEA-Data/Code 2025-022, 497 Pages, 2026/03
JAEA-Data-Code-2025-022.pdf:18.06MB
The Plutonium Fuel Third Development Laboratory of the Nuclear Fuel Cycle Engineering Laboratories has applied for a change of use permit (hereinafter referred to as "license") for plutonium fuel facilities. For the criticality safety design of gloveboxes and equipment/instruments handling mixed oxide (MOX), various criticality calculation codes are used. The most recent employs the 3D Monte Carlo calculation code KENO-V.a embedded in the SCALE 4.4 code system, along with the 27-group ENDF/B-IV neutron cross-section library. SCALE 4.4 was released by the Oak Ridge National Laboratory (ORNL) in the US in 1998, and has now been in use for 27 years. ORNL has continuously improved its functionality, with SCALE 6.3.2 released in 2024. When designing and constructing new MOX fuel facilities, it is desirable to obtain a license using criticality calculation codes based on the latest knowledge. However, it is necessary to verify that these codes have sufficient reliability. Therefore, in 2018, benchmark calculations were performed using the 252-group ENDF/B-VII.1 neutron cross-section library (v7-252n) for two versions of the criticality calculation sequences KENO-V.a and KENO-VI from SCALE 6.2.3, based on past criticality experimental setups. The estimated critical-limiting multiplication factor was calculated. The results indicate that these codes can be used with sufficient confidence for criticality safety design of MOX fuel facilities.
Vauchy, R.; Horii, Yuta; Hirooka, Shun; Akashi, Masatoshi; Sunaoshi, Takeo*; Nakamichi, Shinya; Saito, Kosuke
Journal of Nuclear Materials, 616, p.156115_1 - 156115_16, 2025/10
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)Taniguchi, Yoshinori; Urano, Kenta; Mihara, Takeshi; Udagawa, Yutaka; Kakiuchi, Kazuo; Katsuyama, Jinya
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), p.1292 - 1301, 2025/10
Hirooka, Shun; Vauchy, R.; Horii, Yuta; Sunaoshi, Takeo*; Saito, Kosuke
Proceedings of TopFuel 2025; Nuclear Reactor Fuel Performance Conference (Internet), 10 Pages, 2025/10
Reducing the oxygen-to-metal (O/M) ratio in MOX fuels plays an important role in suppressing the corrosion depth in the cladding due to fuel-cladding chemical interaction (FCCI), which is the key to determining the lifetime of fast reactor MOX fuels. Owing to a number of irradiation and post-irradiation experiments, a clear decreasing trend in the corrosion depth with lower O/M ratio in the as-fabricated MOX pellet was reported, whereas a significant redistribution of the O/M ratio in a pellet driven by the radial temperature gradient during irradiation could supply a higher oxygen potential near the pellet periphery where the FCCI should occur. The reduction of the O/M ratio in the MOX pellet fabrication processes can be achieved by heat treatment by taking high temperature, longer time, and lower oxygen partial pressure in the gas into account; the properties governing the reduction are not sufficiently studied. This study demonstrated the variation of O/M ratio in MOX pellets and the in-situ O/M ratio during the heat treatment was analyzed by using a thermogravimeter, which revealed a decreasing behavior during heating and dwell as well as an increasing behavior in the O/M ratio during the cooling step. Furthermore, the redistribution of O/M ratio, analyzed by Sari's model, was discussed to investigate the O/M ratio and the oxygen potential near the pellet periphery which is likely to have a more direct influence on the FCCI than the as-fabricated O/M ratio. By using the recent oxygen potential data on MOX, it is found that the oxygen potential profile in the radial direction is especially drastic near the pellet periphery and is sensitive to the as-fabricated O/M ratio.
Chauvin, N.*; Martin, P.*; Ogata, Takanari*; Calabrese, R.*; Janney, D.*; Hirooka, Shun; Kato, Masato; Staicu, D.*; McClellan, K.*; White, J.*; et al.
NEA/NSC/R(2024)1 (Internet), 289 Pages, 2025/07
no abstracts in English
Hirooka, Shun; Vauchy, R.; Horii, Yuta; Sunaoshi, Takeo*; Saito, Kosuke; Ozawa, Takayuki
Proceedings of Workshop on Fuel Performance Assessment and Behaviour for Liquid Metal Cooled Fast Reactors (Internet), 8 Pages, 2025/07
no abstracts in English
Calabrese, R.*; Hirooka, Shun
Progress in Nuclear Energy, 178, p.105516_1 - 105516_11, 2025/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Thermal creep is one of the key properties of mixed oxide (MOX) fuel for innovative fast reactors. Thermal creep of fuel affects markedly the interaction between the fuel and the cladding. A review of correlations available in the literature is presented. The effect of porosity, plutonium concentration, and stoichiometry are discussed also in the light of recent numerical results. Our analysis pointed out some inconsistencies concerning the modelling of the effect of porosity on diffusional creep and a re-evaluation of the effect of plutonium concentration. The discussion suggested that Evans's findings on the effect of stoichiometry should be better assessed as well as the level of increase in creep moving towards stoichiometry. Typical operating conditions of fast breeder reactors confirmed the need for an extension of porosity and temperature correlations' domains. Besides this, a new correlation based on a separate-effect approach has been proposed for fuel performance codes.
Vauchy, R.; Hirooka, Shun; Horii, Yuta; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Yamada, Tadahisa*; Tamura, Tetsuya*; Murakami, Tatsutoshi
Journal of Nuclear Materials, 599, p.155233_1 - 155233_11, 2024/10
Times Cited Count：2 Percentile：49.81(Materials Science, Multidisciplinary)The fluorite exsolution/recombination in U
Pu
O
(y = 0.30 and 0.45) and PuO was investigated using differential scanning calorimetry. The results are in relatively good agreement with the literature data, except for plutonia. Our values indicate that the critical temperature of the miscibility gap in Pu-O is 30
50 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U
0Pu
O, U
Pu
O, and PuO dioxides.
Takeuchi, Masayuki; Sano, Yuichi; Sato, Takehiko
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
no abstracts in English
Pu
O
powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Times Cited Count：1 Percentile：27.40(Materials Science, Multidisciplinary)Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：15.37(Nuclear Science & Technology)Kato, Masato; Oki, Takumi; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.; Ozawa, Takayuki; Uwaba, Tomoyuki; Ikusawa, Yoshihisa; Nakamura, Hiroki; Machida, Masahiko
Journal of the American Ceramic Society, 107(5), p.2998 - 3011, 2024/05
Times Cited Count：8 Percentile：47.23(Materials Science, Ceramics)Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：2 Percentile：31.60(Nuclear Science & Technology)
Pu
Am
O
uranium-plutonium-americium mixed oxides at 1573, 1773, and 1873 KVauchy, R.; Sunaoshi, Takeo*; Hirooka, Shun; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Materials, 580, p.154416_1 - 154416_11, 2023/07
Times Cited Count：10 Percentile：88.86(Materials Science, Multidisciplinary)Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji
Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06
Times Cited Count：1 Percentile：10.76(Nuclear Science & Technology)Vauchy, R.; Hirooka, Shun; Matsumoto, Taku; Kato, Masato
Frontiers in Nuclear Engineering (Internet), 1, p.1060218_1 - 1060218_18, 2022/12
PuAm
)O (z = 0.05, 0.10, and 0.15)Yokoyama, Keisuke; Watanabe, Masashi; Tokoro, Daishiro*; Sugimoto, Masatoshi*; Morimoto, Kyoichi; Kato, Masato; Hino, Tetsushi*
Nuclear Materials and Energy (Internet), 31, p.101156_1 - 101156_7, 2022/06
Times Cited Count：6 Percentile：58.85(Nuclear Science & Technology)In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this study, The thermal conductivities of near stoichiometric (UPuAm)O solid solutions(z = 0.05, 0.10, and 0.15) have been measured between room temperature (RT) and 1473 K. The thermal conductivities decreased with increasing Am content and satisfied the classical phonon transport model ((A+BT)
) up to about 1473 K. A values increased linearly with increasing Am content because the change in ionic radius affects the conduction of the phonon due to the solid solution in U
and Am
. B values were independent of Am content.
Hirooka, Shun; Yokoyama, Keisuke; Kato, Masato
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Property studies on Am/Np-bearing MOX were carried out and how the properties influences on the irradiation behaviors was discussed. Both Am and Np inclusions increase the oxygen potential of MOX. Inter-diffusion coefficients obtained by using diffusion couple technique indicate that the inter-diffusion coefficient is larger in the order of U-Am, U-Pu and U-Np. Also, the inter-diffusion coefficients were evaluated to be larger at the O/M = 2 than those of O/M 
2 by several orders. The increase of oxygen potential with Am/Np leads to higher vapor pressure of UO
and the acceleration of the pore migration along temperature gradient during irradiation. The redistributions of actinide elements were also considered with the relationship of the pore migration and diffusion in solid state. Thus, the obtained inter-diffusion coefficients directly influence on the redistribution rate. The obtained properties were modelled and can be installed in a fuel irradiation simulation code.
Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Tasaki, Yudai; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 59(3), p.382 - 394, 2022/03
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)