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Hayashizaki, Kohei; Hirooka, Shun; Yamada, Tadahisa*; Sunaoshi, Takeo*; Murakami, Tatsutoshi; Saito, Kosuke
Ceramics (Internet), 8(1), p.24_1 - 24_12, 2025/03
Hirooka, Shun; Horii, Yuta; Hayashizaki, Kohei; Mohamad, A. B.
Kaku Nenryo, (60-1), p.17 - 20, 2025/02
no abstracts in English
Vauchy, R.; Hirooka, Shun; Saito, Kosuke
Materials Today Communications (Internet), 41, p.110676_1 - 110676_17, 2024/12
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)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.
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
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)The fluorite exsolution/recombination in U
Pu
O (y = 0.30 and 0.45) and PuO 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
50 K lower than previously reported. Finally, the systematic experimental procedure allowed us refining the locus of the solvus existing in hypostoichiometric U
0Pu
O, U
Pu
O, and PuO dioxides.
Pu
)O
(x = 0, 0.18, 0.45, and 1) and analysis of heat capacityHirooka, Shun; Morimoto, Kyoichi; Matsumoto, Taku; Ogasawara, Masahiro*; Kato, Masato; Murakami, Tatsutoshi
Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)no abstracts in English
Pu
O powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Times Cited Count：1 Percentile：68.64(Materials Science, Multidisciplinary) compoundsFrazer, 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
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)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.
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
Times Cited Count：3 Percentile：25.74(Materials Science, Ceramics)
mixed oxidesVauchy, 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
Times Cited Count：6 Percentile：88.73(Materials Science, Multidisciplinary)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
Times Cited Count：6 Percentile：77.44(Materials Science, Multidisciplinary)The thermal conductivities of near-stoichiometric (U,Pu,Am)O doped with NdO
/SmO, 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)
. The dependences of the coefficients A and B on the Nd/Sm content (C
and C
, respectively) are evaluated as: A(mK/W)=1.70 
10
+ 0.93C + 1.20C, B(m/W)=2.39 10
.
Vauchy, R.; Hirooka, Shun; Murakami, Tatsutoshi
Materialia, 32, p.101943_1 - 101943_8, 2023/12
Vauchy, R.; Hirooka, Shun; Murakami, Tatsutoshi
Materialia, 32, p.101934_1 - 101934_12, 2023/12
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
Times Cited Count：5 Percentile：84.10(Nuclear Science & Technology)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.
Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Murakami, Tatsutoshi
Journal of Nuclear Materials, 584, p.154576_1 - 154576_11, 2023/10
Times Cited Count：6 Percentile：84.10(Materials Science, Multidisciplinary)
Pu
Am
O uranium-plutonium-americium mixed oxides at 1573, 1773, and 1873 KVauchy, R.; Sunaoshi, Takeo*; Hirooka, Shun; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato
Journal of Nuclear Materials, 580, p.154416_1 - 154416_11, 2023/07
Times Cited Count：9 Percentile：94.05(Materials Science, Multidisciplinary)Kato, Masato; Nakamichi, Shinya; Hirooka, Shun; Watanabe, Masashi; Murakami, Tatsutoshi; Ishii, Katsunori
Nihon Genshiryoku Gakkai Wabun Rombunshi (Internet), 22(2), p.51 - 58, 2023/04
Uranium and Plutonium mixed oxide (MOX) pellets used as fast reactor fuels have been produced from several raw materials by mechanical blending method through processes of ball milling, additive blending, granulation, pressing, sintering and so on. It is essential to control the pellet density which is one of the important fuel specifications, but it is difficult to understand relationships among many parameters in the production. Database for MOX production was prepared from production results in Japan, and input data of eighteen types were chosen from production process and made a data set. Machine learning model to predict sintered density of MOX pellet was derived by gradient boosting regressor, and represented the measured sintered density with coefficient of determination of R
=0.996
Hirooka, Shun; Nakamichi, Shinya; Matsumoto, Taku; Tsuchimochi, Ryota; Murakami, Tatsutoshi
Frontiers in Nuclear Engineering (Internet), 2, p.1119567_1 - 1119567_7, 2023/03
Storage of plutonium (Pu)-containing materials requires extremely strict attention in terms of physical safety and material accounting. Despite the emphasized importance of storage management, only a few reports are available in the public, e.g., experience in PuO storage in the UK and safety standards in the storage of Pu-containing materials in the US. Japan also stores more U-Pu mixed oxide (MOX) mostly in powder form. Adopting an appropriate storage management is necessary depending on the characteristics of MOX items such as raw powder obtained by reprocessing of spent Light Water Reactor fuels, research and development on the remains of fuel fabrication, which can contain organic materials, and dry-recycled powder during fuel fabrication. Stagnation in fuel fabrications and experience in degradation of MOX containers during extended period of storage have led to the review of the storage method in the Plutonium Fuel Development Center in Japan Atomic Energy Agency. The present work discusses the various nuclear materials, storage methods, experience in degradation of containers that occur during storage, and strategies for future long-term storage.
Vauchy, R.; Hirooka, Shun; Watanabe, Masashi; Kato, Masato
Scientific Reports (Internet), 13, p.2217_1 - 2217_8, 2023/02
Times Cited Count：10 Percentile：78.43(Multidisciplinary Sciences) ceramicsVauchy, R.; Hirooka, Shun; Watanabe, Masashi; Yokoyama, Keisuke; Sunaoshi, Takeo*; Yamada, Tadahisa*; Nakamichi, Shinya; Murakami, Tatsutoshi
Ceramics International, 49(2), p.3058 - 3065, 2023/01
Times Cited Count：11 Percentile：67.57(Materials Science, Ceramics), ThO, UO, PuO, and (U, Pu)OKato, Masato; Watanabe, Masashi; Hirooka, Shun; Vauchy, R.
Frontiers in Nuclear Engineering (Internet), 1, p.1081473_1 - 1081473_10, 2023/01
