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Journal Articles

Oxygen chemical diffusion in fast neutron reactor fuel U$$_{0.698}$$Pu$$_{0.289}$$Am$$_{0.013}$$O$$_{2-x}$$

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

Proceedings of 34th Nuclear Energy for New Europe (NENE2025) , p.232 - 238, 2026/01

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：1 Percentile：41.98(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

None

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

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

no abstracts in English

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：36.47(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

Thermal conductivity measurement of uranium-plutonium mixed oxide doped with Nd/Sm as simulated fission products

Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Materials, 588, p.154799_1 - 154799_20, 2024/01

https://doi.org/10.1016/j.jnucmat.2023.154799
 Times Cited Count：8 Percentile：76.12(Materials Science, Multidisciplinary)

The thermal conductivities of near-stoichiometric (U,Pu,Am)O$$_{2}$$ doped with Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$, which is major fission product (FP) generated by a uranium-plutonium mixed oxides (MOX) fuel irradiation, as simulated fission products are evaluated at 1073-1673 K. The thermal conductivities are calculated from the thermal diffusivities that are measured using the laser flash method. To evaluate the thermal conductivity from a homogeneity viewpoint of Nd/Sm cations in MOX, the specimens with different homogeneity of Nd/Sm are prepared using two kinds of powder made by ball-mill and fusion methods. A homogeneous Nd/Sm distribution decreases the thermal conductivity of MOX with increasing Nd/Sm content, whereas heterogeneous Nd/Sm has no influence. The effect of Nd/Sm on the thermal conductivity is studied using the classical phonon transport model (A+BT)$$^{-1}$$. The dependences of the coefficients A and B on the Nd/Sm content (C$$_{Nd}$$ and C$$_{Sm}$$, respectively) are evaluated as: A(mK/W)=1.70 $$times$$ 10$$^{-2}$$ + 0.93C$$_{Nd}$$ + 1.20C$$_{Sm}$$, B(m/W)=2.39 $$times$$ 10$$^{-4}$$.

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：57.44(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.

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

Development of visual inspection technology of pellets using machine learning, 3; Consideration of an algorithm to identify two types of defects

Goto, Kenta; Hirooka, Shun; Horii, Yuta; Nakamichi, Shinya; Saito, Kosuke; Ichige, Hidekazu; Ono, Takanori; Yamamoto, Kazuya; Hatanaka, Nobuhiro; Murakami, Tatsutoshi

no journal, , 

no abstracts in English

Oral presentation

Thermal conductivity of MOX with simulated fission products

Horii, Yuta; Hirooka, Shun; Kato, Masato; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*

no journal, , 

The low-decontaminated fuel which contains significant amount of fission products (FPs) has been investigated as a fuel for the advanced fast reactor cycle. In this cycle, it is expected to reduce reprocessing cost and strengthen nuclear proliferation resistance of recovered plutonium accompanying high radiation dose FPs. However, FPs could affect on thermal properties of MOX. In particular, thermal conductivity is one of the most important properties for fuel design. Many studies have been reported on the thermal conductivity of MOX that evaluated the effect of plutonium and minor actinide (Am, Np) contents. However, the number of studies on the thermal conductivity of MOX containing FPs are limited; only Nd, Eu and Zr. In this study, thermal conductivity of MOX including Nd$$_{2}$$O$$_{3}$$ and Sm$$_{2}$$O$$_{3}$$, which are expected to be the main FPs remaining as solid solutions in low-decontaminated fuel, was evaluated by measuring thermal diffusivity of sintered pellet. Nd$$_{2}$$O$$_{3}$$ powder and Sm$$_{2}$$O$$_{3}$$ powder was added into raw MOX powder of 20% plutonium content and they were milled by ball-mill. Sintered pellets were obtained by pressing and sintering this powder. The thermal diffusivities were measured by a laser flash method from 973 K to 1673 K at every 100 K. The thermal conductivities up to 1400K are expressed by classical phonon transport model as $$lambda$$=(A+BT)$$^{-1}$$ where A(mK/W)=2.0$$times$$10$$^{-3}$$+4.8$$times$$10$$^{-1}$$C$$_{Nd,Sm}$$ and B(m/W)=2.5$$times$$10$$^{-4}$$. This means the thermal conductivity decreased with enhancement of phonon scattering by adding Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$, similar to the previous studies.

Oral presentation

Demonstration research on fast reactor recycling using low decontaminated MA-bearing MOX fuels, 9; Microstructure and thermal conductivity of MOX with simulated FPs

Horii, Yuta; Hirooka, Shun; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Yamada, Tadahisa*; Furusawa, Naoya*; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

no journal, , 

The low-decontaminated fuel which contains significant amount of fission products (FPs) has been investigated as a fuel for the advanced fast reactor cycle. In this cycle, it is expected to reduce reprocessing cost and strengthen nuclear proliferation resistance of recovered plutonium accompanying high radiation dose FPs. As part of studies on physical properties of low-decontaminated fuel pellets, Nd$$_{2}$$O$$_{3}$$/Sm$$_{2}$$O$$_{3}$$ powders were added to the MOX raw powder as simulated fission products (FPs). The effects of simulated FPs on thermal conductivity were evaluated focusing on the microstructural homogeneities of simulated FPs. From the results of thermal diffusivity measurement and the EPMA mapping, the homogeneous simulated FPs decreased the thermal conductivity of the MOX.

Oral presentation

Development of simplified pelletizing process for fast reactor MOX fuels and demonstration experiment, 4; Development of visual inspection technology of pellets using machine learning

Hirooka, Shun; Horii, Yuta; Segawa, Tomoomi; Kawaguchi, Koichi; Ishii, Katsunori; Makino, Takayoshi; Okumura, Kazuyuki

no journal, , 

no abstracts in English

Oral presentation

Evaluation of burn-up effect on MOX fuel thermal conductivity, 1; Thermal conductivity evaluation of simulated FP-doped MOX

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Uno, Hiroki*; Ogasawara, Masahiro*; Yamada, Tadahisa*; Murakami, Tatsutoshi

no journal, , 

no abstracts in English

Oral presentation

Evaluation of burn-up effect on MOX fuel thermal conductivity, 2; Investigation of thermal conductivity model considering burn-up

Hirooka, Shun; Horii, Yuta; Murakami, Tatsutoshi

no journal, , 

no abstracts in English

Oral presentation

Phase separation/recombination in hypostoichiometric uranium-plutonium mixed oxide U$$_{1-y}$$PuyO$$_{2-x}$$ (y= 0.30; 0.45; and 1.00) using DSC

Vauchy, R.; Ogasawara, Masahiro*; Yamada, Tadahisa*; Sunaoshi, Takeo*; Tamura, Tetsuya*; Horii, Yuta; Hirooka, Shun; Murakami, Tatsutoshi

no journal, , 

The phase separation/recombination occurring in hypostoichiometric uranium-plutonium mixed oxides is important for technological applications, from manufacturing to irradiation behavior. The systematic cross-experiment procedure allowed refining the limits of the solvus existing in hypostoichiometric U$$_{1-y}$$PuyO$$_{2-x}$$ and PuO$$_{2-x}$$ dioxides to better understand the redox behavior of MOX fuel for fast reactors during manufacturing and storage.

Oral presentation

Pu-content dependence of MOX oxygen potential

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

no journal, , 

no abstracts in English

Oral presentation

None

Horii, Yuta; Hirooka, Shun; Vauchy, R.; Uno, Hiroki*; Tamura, Tetsuya*; Sunaoshi, Takeo*; Saito, Kosuke

no journal, , 

no abstracts in English

Oral presentation

Demonstration research on fast reactor recycling using low decontaminated MA-bearing MOX fuels, 2; Sintering test of MOX with simulated FP (Nd$$_{2}$$O$$_{3}$$, Sm$$_{2}$$O$$_{3}$$)

Horii, Yuta; Kato, Masato; Hirooka, Shun; Uno, Hiroki*; Tamura, Tetsuya*; Sunaoshi, Takeo*

no journal, , 

no abstracts in English

Oral presentation

Development of visual inspection technology of pellets using machine learning, 2; Demonstration experiment using MOX pellets and consideration for introduction to production lines

Goto, Kenta; Hirooka, Shun; Horii, Yuta; Nakamichi, Shinya; Murakami, Tatsutoshi; Shibanuma, Kimikazu; Ono, Takanori; Yamamoto, Kazuya; Hatanaka, Nobuhiro; Okumura, Kazuyuki

no journal, , 

no abstracts in English

29 (Records 1-20 displayed on this page)