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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

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

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

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

Journal Articles

Estimation of the activity median aerodynamic diameter of plutonium particles using image analysis

Takasaki, Koji; Yasumune, Takashi; Yamaguchi, Yukako; Hashimoto, Makoto; Maeda, Koji; Kato, Masato

Journal of Nuclear Science and Technology, 60(11), p.1437 - 1446, 2023/11

The aerodynamic radioactive median diameter (AMAD) is necessary information to assess the internal exposure. On June 6, 2017, at a plutonium handling facility in Oarai site of Japan Atomic Energy Agency (JAEA), during the inspection work of a storage container that contains nuclear fuel materials, accidental contamination occurred and five workers inhaled radioactive materials including plutonium. Some smear papers and an air sampling filter were measured with the imaging plate, and we conservatively estimated minimum AMADs for two cases, plutonium nitrate and plutonium dioxide. As a result of AMAD estimation, even excluding a giant particle of a smear sample, the minimum AMADs of plutonium nitrate from smear papers were 4.3 - 11.3 $$mu$$m and those of plutonium dioxide were 5.6 - 14.1 $$mu$$m. Also, the minimum AMAD of plutonium nitrate from an air sampling filter was 3.0 $$mu$$m and that of plutonium dioxide was 3.9 $$mu$$m.

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

 Times Cited Count:1 Percentile:96.85(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

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

 Times Cited Count:1 Percentile:96.85(Materials Science, Multidisciplinary)

Journal Articles

Machine learning sintering density prediction model for MOX fuel pellet

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$$^{2}$$=0.996

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

 Times Cited Count:1 Percentile:93.17(Multidisciplinary Sciences)

Journal Articles

Oxygen potential, oxygen diffusion, and defect equilibria in UO$$_{2 pm x}$$

Watanabe, Masashi; Kato, Masato

Frontiers in Nuclear Engineering (Internet), 1, p.1082324_1 - 1082324_9, 2023/01

Since the oxygen potential and the oxygen coefficient of UO$$_{2}$$ have a significant impact on fuel performance, many experimental data have been obtained. However, experimental data of the oxygen potential and the oxygen diffusion coefficient in the high temperature region above 1673 K are very limited. In the present study, we aimed to obtain these data and analyze them by defect chemistry. The oxygen potentials and the oxygen chemical diffusion coefficient of UO$$_{2}$$ were measured by the gas equilibrium method in the near stoichiometric region at temperatures ranging from 1673 to 1873 K. A data set of oxygen potentials was made together with literature data and analyzed by defect chemistry. The oxygen potential of UO$$_{2}$$ was determined as a function of O/U ratio and temperature, and an equation representing the relationship was derived. The oxygen chemical diffusion coefficient values obtained in this study were reasonably close to the literature values. The oxygen partial pressure dependence of the oxygen chemical diffusion coefficients was predicted from the evaluated results of the oxygen potential data, but no clear dependence was observed.

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

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

Journal Articles

Materials science and fuel technologies of uranium and plutonium mixed oxide

Kato, Masato; Machida, Masahiko; Hirooka, Shun; Nakamichi, Shinya; Ikusawa, Yoshihisa; Nakamura, Hiroki; Kobayashi, Keita; Ozawa, Takayuki; Maeda, Koji; Sasaki, Shinji; et al.

Materials Science and Fuel Technologies of Uranium and Plutonium mixed Oxide, 171 Pages, 2022/10

Innovative and advanced nuclear reactors using plutonium fuel has been developed in each country. In order to develop a new nuclear fuel, irradiation tests are indispensable, and it is necessary to demonstrate the performance and safety of nuclear fuels. If we can develop a technology that accurately simulates irradiation behavior as a technology that complements the irradiation test, the cost, time, and labor involved in nuclear fuel research and development will be greatly reduced. And safety and reliability can be significantly improved through simulation of nuclear fuel irradiation behavior. In order to evaluate the performance of nuclear fuel, it is necessary to know the physical and chemical properties of the fuel at high temperatures. And it is indispensable to develop a behavior model that describes various phenomena that occur during irradiation. In previous research and development, empirical methods with fitting parameters have been used in many parts of model development. However, empirical techniques can give very different results in areas where there is no data. Therefore, the purpose of this study is to construct a scientific descriptive model that can extrapolate the basic characteristics of fuel to the composition and temperature, and to develop an irradiation behavior analysis code to which the model is applied.

Journal Articles

Design and actual performance of J-PARC 3 GeV rapid cycling synchrotron for high-intensity operation

Yamamoto, Kazami; Kinsho, Michikazu; Hayashi, Naoki; Saha, P. K.; Tamura, Fumihiko; Yamamoto, Masanobu; Tani, Norio; Takayanagi, Tomohiro; Kamiya, Junichiro; Shobuda, Yoshihiro; et al.

Journal of Nuclear Science and Technology, 59(9), p.1174 - 1205, 2022/09

 Times Cited Count:3 Percentile:80.29(Nuclear Science & Technology)

In the Japan Proton Accelerator Research Complex, the purpose of the 3 GeV rapid cycling synchrotron (RCS) is to accelerate a 1 MW, high-intensity proton beam. To achieve beam operation at a repetition rate of 25 Hz at high intensities, the RCS was elaborately designed. After starting the RCS operation, we carefully verified the validity of its design and made certain improvements to establish a reliable operation at higher power as possible. Consequently, we demonstrated beam operation at a high power, namely, 1 MW. We then summarized the design, actual performance, and improvements of the RCS to achieve a 1 MW beam.

Journal Articles

Measurements of thermal conductivity for near stoichiometric (U$$_{0.7-z}$$Pu$$_{0.3}$$Am$$_{z}$$)O$$_{2}$$ (z = 0.05, 0.10, and 0.15)

Yokoyama, Keisuke; Watanabe, Masashi; Tokoro, Daishiro*; Sugimoto, Masatoshi*; Morimoto, Kyoichi; Kato, Masato; Hino, Tetsushi*

Nuclear Materials and Energy (Internet), 31, p.101156_1 - 101156_7, 2022/06

 Times Cited Count:0 Percentile:56.43(Nuclear Science & Technology)

In current nuclear fuel cycle systems, to reduce the amount of high-level radioactive waste, minor actinides (MAs) bearing MOX fuel is one option for burning MAs using fast reactor. However, the effects of Am content in fuel on thermal conductivity are unclear because there are no experimental data on thermal conductivity of high Am bearing MOX fuel. In this study, The thermal conductivities of near stoichiometric (U$$_{0.7-z}$$Pu$$_{0.3}$$Am$$_{z}$$)O$$_{2}$$ solid solutions(z = 0.05, 0.10, and 0.15) have been measured between room temperature (RT) and 1473 K. The thermal conductivities decreased with increasing Am content and satisfied the classical phonon transport model ((A+BT)$$^{-1}$$) up to about 1473 K. A values increased linearly with increasing Am content because the change in ionic radius affects the conduction of the phonon due to the solid solution in U$$^{5+}$$ and Am$$^{3+}$$. B values were independent of Am content.

Journal Articles

Recent studies on fuel properties and irradiation behaviors of Am/Np-bearing MOX

Hirooka, Shun; Yokoyama, Keisuke; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04

Property studies on Am/Np-bearing MOX were carried out and how the properties influences on the irradiation behaviors was discussed. Both Am and Np inclusions increase the oxygen potential of MOX. Inter-diffusion coefficients obtained by using diffusion couple technique indicate that the inter-diffusion coefficient is larger in the order of U-Am, U-Pu and U-Np. Also, the inter-diffusion coefficients were evaluated to be larger at the O/M = 2 than those of O/M $$<$$ 2 by several orders. The increase of oxygen potential with Am/Np leads to higher vapor pressure of UO$$_{3}$$ and the acceleration of the pore migration along temperature gradient during irradiation. The redistributions of actinide elements were also considered with the relationship of the pore migration and diffusion in solid state. Thus, the obtained inter-diffusion coefficients directly influence on the redistribution rate. The obtained properties were modelled and can be installed in a fuel irradiation simulation code.

Journal Articles

Advanced reactor experiments for sodium fast reactor fuels (ARES) project; Transient irradiation experiments for metallic and MOX fuels

Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

Journal Articles

Defect equilibria and thermophysical properties of CeO$$_{2-x}$$ based on experimental data and density functional theory calculation result

Watanabe, Masashi; Nakamura, Hiroki; Suzuki, Kiichi; Machida, Masahiko; Kato, Masato

Journal of the American Ceramic Society, 105(3), p.2248 - 2257, 2022/03

 Times Cited Count:1 Percentile:7.37(Materials Science, Ceramics)

Properties of CeO$$_{2}$$ were evaluated by DFT simulation to determine band gap, Frenkel defect formation energy and defect migration energy. Band gap and Frenkel defect formation energy were used to analyze defect equilibria. Oxygen partial pressure dependence of defect equilibria was evaluated based on oxygen potential experimental data and DFT calculation, and a Brouwer diagram was derived. The defect formation energies, including Frenkel defect, electron-hole pair and so on, were determined and used to evaluate the properties, including oxygen diffusion coefficients, electrical conduction, heat capacity and thermal conductivity. Mechanisms of various properties were discussed for a deeper understanding based on defect chemistry, and the relationship among properties were systematically described.

Journal Articles

Development of fuel performance analysis code, BISON for MOX, named Okami; Analyses of pore migration behavior to affect the MA-bearing MOX fuel restructuring

Ozawa, Takayuki; Hirooka, Shun; Kato, Masato; Novascone, S.*; Medvedev, P.*

Journal of Nuclear Materials, 553, p.153038_1 - 153038_16, 2021/09

 Times Cited Count:2 Percentile:58.27(Materials Science, Multidisciplinary)

To evaluate the O/M dependence of pore migration regarding fuel restructuring at the beginning of irradiation, we are developing BISON for MOX in cooperation with INL and have installed pore migration model considering vapor pressure of vapor species and thermal conductivity for MOX. The O/M dependence of fuel restructuring observed in MA-bearing MOX irradiation experiment in Joyo was evaluated by the 2-dimensional analyses. Four MA-bearing MOX pins with different O/M ratio and pellet/cladding gap size were irradiated in Joyo B14 experiment. Remarkable restructuring of stoichiometric MA-bearing MOX fuels was observed in PIE, and could be evaluated by considering the influence of O/M ratio on vapor pressure. Also, a central void assumes to move toward wide-gap side when the pellet eccentricity taking place, but 2-dimentional analyses on pellet transverse section revealed that the central void formation observed in PIE would be inconsistent with a direction of the pellet eccentricity.

Journal Articles

Fuel design and fabrication; Pellet-type fuel

Kato, Masato

Encyclopedia of Nuclear Energy, Vol.2, p.298 - 307, 2021/00

Journal Articles

Oxygen potential measurement of (U,Pu,Am)O$$_{2 pm x}$$ and (U,Pu,Am,Np)O$$_{2 pm x}$$

Hirooka, Shun; Matsumoto, Taku; Kato, Masato; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*

Journal of Nuclear Materials, 542, p.152424_1 - 152424_9, 2020/12

 Times Cited Count:4 Percentile:61.45(Materials Science, Multidisciplinary)

The measurement of oxygen potential was conducted at 1,673, 1,773, and 1,873 K for (U$$_{0.623}$$Pu$$_{0.350}$$Am$$_{0.027}$$)O$$_{2}$$ and at 1,873 and 1,923 K for (U$$_{0.553}$$Pu$$_{0.285}$$Am$$_{0.015}$$Np$$_{0.147}$$)O$$_{2}$$ by using a thermo-gravimeter and an oxygen sensor. Am inclusion in terms of substituting the U significantly increased the oxygen potential. Similarly, the inclusion of Np as a substitute for U increased the oxygen potential; however, the effect was not as large as that with the Pu or Am addition at the same rate. The results were analyzed via defect chemistry and certain defect formations were suggested in the reducing region and the near-stoichiometric region by plotting the relationship between PO$$_{2}$$ and the deviation from the stoichiometry. The equilibrium constants of the defect reactions were arranged to reproduce the experiment such that Am/Np contents were included in the entropy with coefficients fitting the experimental data.

356 (Records 1-20 displayed on this page)