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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

 Times Cited Count:0 Percentile:0.02(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

 Times Cited Count:1 Percentile:89.37(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

Toward long-term storage of nuclear materials in MOX fuels fabrication facility

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$$_{2}$$ 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.

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:85.37(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

 Times Cited Count:4 Percentile:75.97(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

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, p.1 - 11, 2023/00

 Times Cited Count:1 Percentile:89.37(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

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

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

Relative oxygen potential measurements of (U,Pu)O$$_{2}$$ with Pu = 0.45 and 0.68 and related defect formation energy

Hirooka, Shun; Matsumoto, Taku; Sunaoshi, Takeo*; Hino, Tetsushi*

Journal of Nuclear Materials, 558, p.153375_1 - 153375_8, 2022/01

 Times Cited Count:1 Percentile:21.98(Materials Science, Multidisciplinary)

no abstracts in English

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:41.01(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

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:51.58(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.

Journal Articles

Effect of O/M ratio on sintering behavior of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2-x}$$

Nakamichi, Shinya; Hirooka, Shun; Kato, Masato; Sunaoshi, Takeo*; Nelson, A. T.*; McClellan, K. J.*

Journal of Nuclear Materials, 535, p.152188_1 - 152188_8, 2020/07

 Times Cited Count:7 Percentile:72.9(Materials Science, Multidisciplinary)

Oxygen-to-metal ratio (O/M) of uranium and plutonium mixed oxide depends on its oxygen partial pressure. To attain the desirable microstructure and O/M ratio of sintered pellets, it is important to investigate the relation between the sintering behavior and the atmosphere of sintering process. In this study, sintering behavior of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2}$$ and (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{1.99}$$ in controlled po$$_{2}$$ atmosphere were investigated. It was found activation energy of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{1.99}$$ was higher than that of (Pu$$_{0.3}$$U$$_{0.7}$$)O$$_{2}$$. On the other hand, it was observed grain growth during sintering was suppressed in hypo-stoichiometric composition.

Journal Articles

Physical properties of non-stoichiometric (U,Pu)O$$_{2}$$

Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Morimoto, Kyoichi; Kato, Masato

2018 GIF Symposium Proceedings (Internet), p.315 - 320, 2020/05

Recently, a research group studying at Plutonium Fuel Development Facility (PFDF) in Japan Atomic Energy Agency has systematically measured vast amounts of physical properties in the non-stoichiometric (U, Pu)O$$_{2}$$. Lattice parameter, elastic modulus, thermal expansion, oxygen potential, oxygen chemical diffusion coefficient and thermal conductivity were successfully measured as function of Pu content, O/M ratio and temperature, and the effects of Pu content and O/M ratio on their physical properties were evaluated. In this work, those experimental data are reviewed, and latest experimental data set on the non-stoichiometric (U, Pu)O$$_{2}$$ are presented. The data set would be available in development of a fuel performance code.

Journal Articles

Oxygen potential and self-irradiation effects on fuel temperature in Am-MOX

Ikusawa, Yoshihisa; Hirooka, Shun; Uno, Masayoshi*

2018 GIF Symposium Proceedings (Internet), p.321 - 327, 2020/05

Research and development of Minor actinides (MAs) bearing MOX fuel for fast reactor has been proceeding from the viewpoint of reducing radioactive waste. In order to develop, MA bearing MOX, it is indispensable to clarify the influence of MA addition on irradiation behavior. The addition of Americium (Am) to MOX affects vapor pressure and thermal conductivity, which are important properties from the perspective of evaluating fuel temperature. This is because vapor pressure affects fuel restructuring, and thermal conductivity affects fuel temperature distribution. Focusing on these physical properties, this study evaluates the influence of Am on fuel temperature using irradiation behavior analysis code to contribute to the development of MA-bearing MOX fuel. An increase in Am content decreases the thermal conductivity and increases the oxygen potential of oxide fuel. Because vapor pressure increases with increasing Am content, pore migration is accelerated, and the central void diameter increases with increasing Am content. As a result, after formation of the central void, the influence of Am content on the fuel center temperature is mild. Alpha particles generated by radioactive decay of transuranium elements cause lattice defects in the oxide fuel pellets. It is well known that this phenomenon, which is called self-irradiation, affects thermal conductivity. Since americium is the typical alpha radioactive nucleus, to evaluate fuel temperature of Am-MOX is necessary to take account of the influence of self-irradiation damage on thermal conductivity. Self-irradiation decreases thermal conductivity, and as the Am content increases, the rate of decrease in thermal conductivity is accelerated. Because it recovers with temperature rise, the decrease in thermal conductivity due to self-irradiation damage has very little effect on fuel center temperature. These results suggest that Am-MOX fuel could be irradiated under the same conditions as conventional MOX fuel.

Journal Articles

Modeling and simulation of redistribution of oxygen-to-metal ratio in MOX

Hirooka, Shun; Kato, Masato; Watanabe, Masashi

Transactions of the American Nuclear Society, 118, p.1624 - 1626, 2018/06

This study suggested the time development of oxygen-to-metal ratio (O/M) redistribution model with oxygen-related properties in MOX. Irradiation simulation including the suggested O/M redistribution and pore migration with vaporization-condensation model which bares density redistribution was demonstrated. The simulation results showed that O/M redistribution proceeded at lower temperature than density redistribution, which indicated that oxygen diffusion got influential at lower temperature than vaporization-condensation of MOX. Another find was that O/M redistribution was very slow at the surface because temperature kept low. However, near the surface (inside from the surface) where the temperature exceeded 1000 K, O/M redistribution was rather recognizable with oxygen flown from inner region to the near-surface. The results will be evaluated by comparison with post-irradiation examination data.

Journal Articles

Sound speeds in and mechanical properties of (U,Pu)O$$_{2-x}$$

Hirooka, Shun; Kato, Masato

Journal of Nuclear Science and Technology, 55(3), p.356 - 362, 2018/03

 Times Cited Count:6 Percentile:56.14(Nuclear Science & Technology)

The sound speeds of longitudinal and transverse waves in the uranium-plutonium mixed oxide (MOX) pellets were measured as functions of porosity, oxygen-to-metal ratio (O/M) and plutonium content. The effect of each parameter was well fitted by a linear function and the equations were obtained to calculate the sound speeds. Mechanical properties were evaluated with the sound speeds and the result of Young's modulus showed that porosity was the most important factor to decrease Young's modulus. Temperature dependence on Young's modulus was also evaluated with previously reported thermal expansion. Decrease of Young's modules with increasing temperature was in good agreement with available literature.

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