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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：0 Percentile：0.00(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 30$$sim$$50 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：2 Percentile：86.32(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：68.64(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：3 Percentile：25.74(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：6 Percentile：88.73(Materials Science, Multidisciplinary)

Journal Articles

First foreigner handling plutonium in a Japanese MOX facility

Vauchy, R.

Kaku Nenryo, (59), 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：84.10(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.

Journal Articles

Lattice parameters of fluorite-structured uranium-americium mixed oxides

Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Murakami, Tatsutoshi

Journal of Nuclear Materials, 584, p.154576_1 - 154576_11, 2023/10

https://doi.org/10.1016/j.jnucmat.2023.154576
 Times Cited Count：6 Percentile：84.10(Materials Science, Multidisciplinary)

Journal Articles

Oxygen potential of neodymium-doped U$$_{0.817}$$Pu$$_{0.180}$$Am$$_{0.003}$$O$$_{2 pm x}$$ uranium-plutonium-americium mixed oxides at 1573, 1773, and 1873 K

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

Journal of Nuclear Materials, 580, p.154416_1 - 154416_11, 2023/07

https://doi.org/10.1016/j.jnucmat.2023.154416
 Times Cited Count：9 Percentile：94.05(Materials Science, Multidisciplinary)

Journal Articles

Breaking the hard-sphere model with fluorite and antifluorite solid solutions

Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Kato, Masato

Scientific Reports (Internet), 13, p.2217_1 - 2217_8, 2023/02

https://doi.org/10.1038/s41598-023-29326-0
 Times Cited Count：10 Percentile：78.43(Multidisciplinary Sciences)

Journal Articles

Liquid phase sintering of alumina-silica co-doped cerium dioxide CeO$$_{2}$$ ceramics

Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Sunaoshi, Takeo*; Yamada, Tadahisa*; Nakamichi, Shinya; Murakami, Tatsutoshi

Ceramics International, 49(2), p.3058 - 3065, 2023/01

https://doi.org/10.1016/j.ceramint.2022.09.295
 Times Cited Count：11 Percentile：67.57(Materials Science, Ceramics)

Journal Articles

Oxygen diffusion in the fluorite-type oxides CeO$$_{2}$$, ThO$$_{2}$$, UO$$_{2}$$, PuO$$_{2}$$, and (U, Pu)O$$_{2}$$

Kato, Masato; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.

Frontiers in Nuclear Engineering (Internet), 1, p.1081473_1 - 1081473_10, 2023/01

https://doi.org/10.3389/fnuen.2022.1081473

Journal Articles

Cation interdiffusion in uranium-plutonium mixed oxide fuels; Where are we now?

Vauchy, R.; Hirooka, Shun; Matsumoto, Taku; Kato, Masato

Frontiers in Nuclear Engineering (Internet), 1, p.1060218_1 - 1060218_18, 2022/12

https://doi.org/10.3389/fnuen.2022.1060218

Oral presentation

Study on controlling low O/M ratio in MOX pellet, 1; Variation of O/M ratio during whole heat treatment

Hirooka, Shun; Vauchy, R.; Horii, Yuta; Akashi, Masatoshi; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

no abstracts in English

Oral presentation

Manufacturing and property measurements of homogeneous simulated FP (Nd/Sm/Gd/Zr)-doped MOX

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Hayashizaki, Kohei; Uno, Hiroki*; Tamura, Tetsuya*; Sunaoshi, Takeo*; Owada, Hideaki*; Yamada, Tadahisa*; Murakami, Tatsutoshi

no journal, , 

Fission products (FPs), which are generated and stored in fuel matrix by irradiating nuclear fuels, affect thermo-physical fuel properties. To improve accuracy of computer simulation of irradiation behaviors, studies on the fuel properties containing FPs are needed. However, only a limited number of studies on the irradiated fuel properties, especially MOX fuels, have been reported in the world due to difficulties in handling of the irradiated fuels. Moreover, the effect of individual FP cannot be evaluated because many kinds of FPs are stored in the irradiated fuels. Thus, an alternative method should be suggested to easily study the effects of FPs on the fuel properties. In this study, fuel properties were measured to evaluate the effects of FPs by using simulated FP-doped MOX specimens instead of a real irradiated fuel. For the measurement, the homogeneity of FP in a specimen is also important, as well as uranium and plutonium. To obtain homogeneous specimens, re-grinding and re-sintering processes were repeated and the improvement was confirmed by EPMA and XRD at each set of the process. A specimen with suitable homogeneity for measurement was prepared by repeating the series of processes three times. Sm$$_{2}$$O$$_{3}$$, Gd$$_{2}$$O$$_{3}$$ and ZrO$$_{2}$$, which are major and soluble FPs in irradiated MOX fuels, were selected as simulated FPs. The effect of individual FP on the properties, such as thermal conductivity and thermal expansion, was evaluated on the specimens. In addition, Nd$$_{2}$$O$$_{3}$$, Sm$$_{2}$$O$$_{3}$$ and Gd$$_{2}$$O$$_{3}$$ co-doped MOX was also prepared to compare the influence of containing multiple lanthanides.

Oral presentation

MOX fuel technologies for next-generation nuclear reactors, 1; Evaluation of electronic defect contribution in heat capacity and thermal conductivity

Kato, Masato; Hirooka, Shun; Watanabe, Masashi; Vauchy, R.; Oki, Takumi

no journal, , 

no abstracts in English

Oral presentation

Oxygen potential measurements of Nd doped (U,Pu,Am)O$$_{2}$$ mixed oxides

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

no journal, , 

32 (Records 1-20 displayed on this page)