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Journal Articles

A Thermodynamic study of fluorite/bixbyite equilibria in the UO $_{2}$ -PuO $_{1.5}$ -PuO $_{2}$ system

Vinograd, V. L.*; Vauchy, R.

Journal of Nuclear Materials, 619, p.156244_1 - 156244_16, 2026/01

https://doi.org/10.1016/j.jnucmat.2025.156244

Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)

Thermodynamic properties of (U $_{1-z}$ Pu $_{z}$ )O $_{2+delta}$ fluorite (FCC) and (U $_{1-z}$ Pu $_{z}$ )O $_{2+delta}$ bixbyite (BCC) in the UO $_{2}$ -PuO $_{1.5}$ -PuO $_{2}$ system are assessed by considering phase equilibrium constraints and data on the variation of oxygen to metal ratio (O/M) as a function of the chemical potential of O $_{2}$ . Thermodynamically, both BCC and FCC are described as ordered solid solutions allowing for a decrease in the configurational entropy of their oxygen/vacancy distributions at the specific values of = -0.5 and = -0.375 (O/M = 1.5 and O/M = 1.625). With this approach, fluorite/bixbyite equilibria in PuO $_{1.5}$ -PuO $_{2}$ and in UO $_{2}$ -PuO $_{1.5}$ -PuO $_{2}$ are reproduced well with little effort. Moreover, we show that a large manifold of experimental data on the UO $_{2}$ -PuO $_{1.5}$ -PuO $_{2}$ system is consistent with the assumption that Pu/(Pu+U) ratios in individual phases remain equal to the total Pu/(Pu+U) ratio in the system, i.e., no inter-phase U/Pu-partitioning occurs.

Journal Articles

Oxygen potential and oxygen diffusion data for guiding the manufacture of MOX fuel for fast neutron reactors

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

https://doi.org/10.1016/j.jnucmat.2025.156115

Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)

Journal Articles

Fabrication of low-O/M fast reactor MOX fuel and analysis on its oxygen potential behaviors

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.

Journal Articles

Control and irradiation behaviors of oxygen potential of MOX fuel

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

Journal Articles

Report on the "Plutonium futures; The Science 2024" International Conference (Charleston, South Carolina, USA)

Vauchy, R.

Kaku Nenryo, (60-1), p.9 - 11, 2025/02

Plutonium Futures is historically an international conference focusing on plutonium science. As a result, one of the main topics is naturally military applications (nuclear weapons) but nuclear fuels (oxides, carbides, nitrides, metals) and reprocessing are also now a large part the topics presented at the conference.

Journal Articles

Oxygen potential of plutonium and plutonium-americium dioxides

Vauchy, R.; Hirooka, Shun; Saito, Kosuke

Materials Today Communications (Internet), 41, p.110676_1 - 110676_17, 2024/12

https://doi.org/10.1016/j.mtcomm.2024.110676

Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)

Oxygen potential measurements of PuO $_{2-x}$ reported in the open literature were reviewed and re-interpreted using the defect chemistry model developed by our team. An empirical, easy-to-use, relationship connecting the O/Pu ratio, the equilibrium oxygen potential, and the temperature is proposed based on the interpolation of the experimental data in the 953-2100 K temperature range. The effect of americium on the oxygen potential of PuO $_{2-x}$ is also discussed.

Journal Articles

Uranium-plutonium-oxygen phase diagram; Investigating the solvus of fluorite's exsolution

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

https://doi.org/10.1016/j.jnucmat.2024.155233

Times Cited Count：2 Percentile：49.81(Materials Science, Multidisciplinary)

The fluorite exsolution/recombination in U $_{1-y}$ Pu $_{y}$ O $_{2-x}$ (y = 0.30 and 0.45) and PuO $_{2-x}$ 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 3050 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U $_{0.70}$ 0Pu $_{0.30}$ O $_{2-x}$ , U $_{0.55}$ Pu $_{0.45}$ O $_{2-x}$ , and PuO $_{2-x}$ dioxides.

Journal Articles

Oxygen potential measurement of U $_{0.85}$ Am $_{0.15}$ O $_{2}$ at 1473, 1573, and 1673 K

Watanabe, 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

https://doi.org/10.1016/j.jnucmat.2024.155232

Times Cited Count：3 Percentile：65.33(Materials Science, Multidisciplinary)

Oxygen potential data of U $_{0.85}$ Am $_{0.15}$ O $_{2-x}$ were measured at 1473, 1573, and 1673 K by thermogravimetry. In U $_{1-y}$ An $_{y}$ O $_{2-x}$ , where An stands for Pu or Am, and for a given value of y and Oxygen/Metal ratio, the oxygen potential of U $_{1-y}$ Am $_{y}$ O $_{2-x}$ is higher than that of U $_{1-y}$ Pu $_{y}$ O $_{2-x}$ . The valence of cations in the hypostoichiometric region is similar to that of Nd-doped UO $_{2}$ . At the stoichiometric composition, it is estimated to be Am $^{3+}$ , U $^{4+}$ , and U $^{5+}$ (for charge compensation of Am $^{3+}$ ). 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.

Journal Articles

Sintering behavior analysis of compacted dry recycled U $_{0.7}$ Pu $_{0.3}$ O $_{2}$ powder using master sintering curve theory

Nakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi

Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07

https://doi.org/10.1016/j.jnucmat.2024.155072

Times Cited Count：1 Percentile：27.40(Materials Science, Multidisciplinary)

Journal Articles

A Science-based mixed oxide property model for developing advanced oxide nuclear fuels

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

https://doi.org/10.1111/jace.19609

Times Cited Count：8 Percentile：47.23(Materials Science, Ceramics)

Journal Articles

Uranium-plutonium-americium cation interdiffusion in polycrystalline (U,Pu,Am)O $_{2 pm x}$ mixed oxides

Vauchy, R.; Matsumoto, Taku; Hirooka, Shun; Uno, Hiroki*; Tamura, Tetsuya*; Arima, Tatsumi*; Inagaki, Yaohiro*; Idemitsu, Kazuya*; Nakamura, Hiroki; Machida, Masahiko; et al.

Journal of Nuclear Materials, 588, p.154786_1 - 154786_13, 2024/01

https://doi.org/10.1016/j.jnucmat.2023.154786

Times Cited Count：9 Percentile：86.25(Materials Science, Multidisciplinary)

Journal Articles

First foreigner handling plutonium in a Japanese MOX facility

Vauchy, R.

Kaku Nenryo, (59-1), p.13 - 15, 2023/12

Journal Articles

Ionic radii in fluorites

Vauchy, R.; Hirooka, Shun; Murakami, Tatsutoshi

Materialia, 32, p.101934_1 - 101934_12, 2023/12

https://doi.org/10.1016/j.mtla.2023.101934

Journal Articles

Ionic radii in halites

Vauchy, R.; Hirooka, Shun; Murakami, Tatsutoshi

Materialia, 32, p.101943_1 - 101943_8, 2023/12

https://doi.org/10.1016/j.mtla.2023.101943

Journal Articles

Sintering and microstructural behaviors of mechanically blended Nd/Sm-doped MOX

Hirooka, Shun; Horii, Yuta; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*; Vauchy, R.; Hayashizaki, Kohei; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Science and Technology, 60(11), p.1313 - 1323, 2023/11

https://doi.org/10.1080/00223131.2023.2192725

Times Cited Count：5 Percentile：66.06(Nuclear Science & Technology)

Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. Nd $_{2}$ O $_{3}$ and Sm $_{2}$ O $_{3}$ are chosen as the additive materials to simulate the corresponding soluble fission products dispersed in MOX. Shrinkage curves of the MOX pellets are obtained by dilatometry, which reveal that the sintering temperature is shifted toward a value higher than that of the respective regular MOX. The additives, however, promote grain growth and densification, which can be explained by the effect of oxidized uranium cations covering to a pentavalent state. Ceramography reveals large agglomerates after sintering, and Electron Probe Micro-Analysis confirms that inhomogeneous elemental distribution, whereas XRD reveals a single face-centered cubic phase. Finally, by grinding and re-sintering the specimens, the cation distribution homogeneity is significantly improved, which can simulate spent nuclear fuels with soluble fission products.