JOPSS - Search Results

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

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：75.80(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

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 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

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：51.66(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：5 Percentile：41.38(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：84.29(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：78.63(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：84.29(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：93.25(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：72.61(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：63.47(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

MOX fuel technologies; France vs. Japan

Vauchy, R.

no journal, ,

A comparison of the French and Japanese nuclear fuel cycles is proposed with a focus on MOX fuel production for LWRs and SFRs.

Oral presentation

Experimental measurements of oxygen potential of neodymium doped MOX fuel

Vauchy, R.; Hirooka, Shun; Sunaoshi, Takeo*; Kato, Masato

no journal, ,

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

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Oxygen potential of plutonium and plutonium-americium dioxides

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

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

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

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

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

First foreigner handling plutonium in a Japanese MOX facility

Ionic radii in fluorites

Ionic radii in halites

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

Lattice parameters of fluorite-structured uranium-americium mixed oxides

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

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

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

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

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

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

MOX fuel technologies; France vs. Japan

Experimental measurements of oxygen potential of neodymium doped MOX fuel

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