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

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

None

Hirooka, Shun; Horii, Yuta; Hayashizaki, Kohei; Mohamad, A. B.

Kaku Nenryo, (60-1), p.17 - 20, 2025/02

no abstracts in English

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

Oxygen potential measurements of PuO 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 is also discussed.

Enthalpy measurement on (UPu)O (x = 0, 0.18, 0.45, and 1) and analysis of heat capacity

Hirooka, Shun; Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*; Kato, Masato; Murakami, Tatsutoshi

Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09

https://doi.org/10.1016/j.jnucmat.2024.155188

no abstracts in English

High temperature nanoindentation of (U,Ce)O compounds

Frazer, D.*; Saleh, T. A.*; Matsumoto, Taku; Hirooka, Shun; Kato, Masato; McClellan, K.*; White, J. T.*

Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07

https://doi.org/10.1016/j.nucengdes.2024.113136

Nanoindentation based techniques can be employed on minute volumes of material to measure mechanical properties, including Young's modulus, hardness, and creep stress exponents. In this study, (U,Ce)O solid solutions samples are used to develop elevated temperature nanoindentation and nanoindentation creep testing methods for use on mixed oxide fuels. Nanoindentation testing was performed on 3 separate (Ux-1,Cex)O compounds ranging from x equals 0.1 to 0.3 at up to 800 C: their Young's modulus, hardness, and creep stress exponents were evaluated. The Young's modulus decreases in the expected linear manner while the hardness decreases in the expected exponential manner. The nanoindentation creep experiments at 800 C give stress exponent values, n=4.7-6.9, that suggests dislocation motion as the deformation mechanism.

Uranium-plutonium-americium cation interdiffusion in polycrystalline (U,Pu,Am)O 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

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

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

Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. NdO and SmO 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.

EBR-II MOX fuel characterization enabling ARES Phase I testing

Bess, J. D.*; Chipman, A. S.*; Pope, C. L.*; Jensen, C. B.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato*

Nuclear Science and Engineering, 197(8), p.1845 - 1872, 2023/08

https://doi.org/10.1080/00295639.2022.2156240

Pretransient characterization was performed for the EBR-II MOX fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and US. The continued collaboration will investigate the transient performance of these rods in TREAT at Idaho National Laboratory. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of ~134.4 GWd/t in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination.