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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：3 Percentile：95.99(Nuclear Science & Technology)Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. Nd
O
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.
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：4 Percentile：98.08(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.
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：8 Percentile：75.06(Materials Science, Ceramics)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.
U
)O
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：9 Percentile：75.92(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 (PuU)O and (PuU)O
in controlled po atmosphere were investigated. It was found activation energy of (PuU)O was higher than that of (PuU)O. On the other hand, it was observed grain growth during sintering was suppressed in hypo-stoichiometric composition.
and (U,Pu)ONakamichi, Shinya; Hirooka, Shun; Sunaoshi, Takeo*; Kato, Masato; Nelson, A.*; McClellan, K.*
Transactions of the American Nuclear Society, 113(1), p.617 - 618, 2015/10
Cerium dioxide has been used as a surrogate material for plutonium dioxide. Dorr et al reported the use of hyper-stoichiometric conditions causes the start of shrinkage of (U,Ce)O at low temperature compared with the sintering in reducing atmosphere. However, the precise stoichiometry of the samples investigated was not controlled or otherwise monitored, preventing any quantitative conclusions regarding the similarities or differences between (U,Ce)O and (U,Pu)O. The motivation for the present work is therefore to compare the sintering behavior of MOX and the (U,Ce)O MOX surrogates under controlled atmospheres to assess the role of oxygen defects on densification in both systems.
Kitamura, Akihiro; Nakamichi, Shinya; Kashiro, Kashio
Hoken Butsuri, 47(1), p.66 - 73, 2012/04
Plasma arc cutting is considered to be one of the effective methods to size reduce facility components. This method produces an electrical arc to cut the metal equipment, and during the cutting, a molten metal is ejected in the form of aerosol, which is blown away from the torch and has a potential to spread all over the environment. In order to confirm the safety use of plasma arc cutting for dismantling activity in plutonium fuel fabricating facility, we have performed experiments to find aerosol size distribution and to confirm effectiveness of HEPA filter during the cutting. Results showed that the activity median aerosol diameter was around 6 micrometers, with a geometric standard deviation of 1.9. Also, radiological survey was conducted on the back surface of HEPA filter and confirmed that the penetration of radioactive aerosol was under detection limit.
Pu
)O
at 1773 and 1873 K, and its analysis based on point defect chemistryKato, Masato; Nakamichi, Shinya; Takeuchi, Kentaro; Sunaoshi, Takeo*
CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 35(4), p.623 - 626, 2011/12
Times Cited Count：12 Percentile：54.56(Thermodynamics)Uranium and plutonium mixed oxide (MOX) has been used as fuels of fast reactors. The MOX having fluorite structure is an oxygen nonstoichiometric compound which is stable in hyper- and hypo-stoichiometric composition range. The stoichiometry of MOX significantly affects their properties. So, it is essential to know the relation between stoichiometry and oxygen potential to develop MOX fuels. In this work, the oxygen potentials of (UPu)O were measured at high temperatures of 1773, and 1873K. The measurements were carried out by gas equilibrium method using thermo-gravimetry. Th The oxygen partial pressure was adjusted by controlling the ratio of H/HO in the flowing gas atmosphere, and the oxygen potential was determined. The oxygen potentials were determined as functions of O/M ratio, and temperature. The data at stoichiometric composition were estimated to be -311kJ/mol and -299kJ/mol at 1773K, and 1873K based on point defect model.
Nakamichi, Shinya; Kato, Masato; Tamura, Tetsuya*
CALPHAD; Computer Coupling of Phase Diagrams and Thermochemistry, 35(4), p.648 - 651, 2011/12
Times Cited Count：11 Percentile：52.23(Thermodynamics)Japan Atomic Energy Agency has developed homogeneous MOX fuels containing minor actinides (MAs) such as Am and Np elements. In this work, particular attention is paid to the influence of small MA addition on oxygen potential of MOX fuels. Oxygen potentials of (Am
Pu
U
)O and (Am
Np
Pu
U
)O in the near stoichiometric region were measured at 1473, 1573 and 1623K by thermogravimetry. (AmPuU)O and (AmNpPuU)O were slightly higher than those of (Pu
U
)O. From this work, it was found that oxygen potential of MOX is slightly increased by small Am addition and influence of Np addition on oxygen potential is not significant.
Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki
Proceedings of 10th OECD/NEA Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation (CD-ROM), p.201 - 209, 2010/00
Japan Atomic Energy Agency has developed homogeneous MOX fuel containing minor actinide (MA) elements such as Np and Am. To measure physical properties of the fuel is essential for its development, because their data are needed to evaluate irradiation behavior. In this report, the physical properties, melting temperature, thermal conductivity, lattice parameter, oxygen potential and phase separation behavior, were reviewed, and effect of MA content was discussed.
systemNakamichi, Shinya; Kato, Masato; Sunaoshi, Takeo*; Uchida, Teppei; Morimoto, Kyoichi; Kashimura, Motoaki; Kihara, Yoshiyuki
Journal of Nuclear Materials, 389(1), p.191 - 196, 2009/05
Times Cited Count：2 Percentile：17.7(Materials Science, Multidisciplinary)Japan Atomic Energy Agency researchers have developed mixed oxide (MOX) fuels containing minor actinides (MA). These fuels were irradiated for ten minutes in the FBR Joyo in some short-term irradiation tests. The Si-condensed phases were observed at the center of the pellets in the post irradiation examination. Si impurities came to be mixed into the raw materials in the ball milling process, because Si rubber was used as the lining of the milling pot. Content of Si in the pellets was within the specification of the fuel. It is important to investigate the Si state in MOX at high temperatures like the reactor operating temperature of the fuel to evaluate irradiation behavior. In the present work, MOX specimens with mixed SiO impurity were prepared. The ratio of MOX to SiO was controlled at a mol fraction of 3 to 1. The specimens were first heated at 1973K in atmospheres of three different oxygen partial pressures to adjust the O/M ratio. Then these specimens were sealed in a tungsten capsule, and heated at 2273K or 2673K. Compounds consisting of Pu and Si were observed at grain boundaries of the MOX matrix in specimens after heat treatment. These compounds were not observed in grain interior and MOX matrix was not affected significantly by Si impurity. These compounds tended to form in specimens with low O/M ratio and in specimens heated at higher temperatures.
Kato, Masato; Morimoto, Kyoichi; Nakamichi, Shinya; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(4), p.420 - 428, 2008/12
The melting temperatures of MOX for fast reactor fuel were investigated as functions of Pu content, Am content and oxygen-to-metal (O/M) ratio using thermal arrest technique. Rhenium inner was used for the measurement to prevent the reaction between the sample and capsule materials. The solidus temperatures decreased with increasing Pu and Am content and increased with decreasing O/M ratio. It is considered that the maximum temperature in U-Pu-O system varies in hypostoichiometric composition region. The melting temperatures were evaluated by ideal solid solution model in UO-PuO-AmO-PuO
system, and the model was derived for calculating solidus and liquidus temperature. The derived model reproduced the experimental data with 
25 K.
Kato, Masato; Nakamichi, Shinya; Takano, Tatsuo
Proceedings of International Conference on Advanced Nuclear Fuel Cycles and Systems (Global 2007) (CD-ROM), p.916 - 920, 2007/09
no abstracts in English
Am
U
)O
Nakamichi, Shinya; Kato, Masato; Morimoto, Kyoichi; Sugata, Hiromasa*; Kashimura, Motoaki; Abe, Tomoyuki
Transactions of the American Nuclear Society, 96(1), p.191 - 192, 2007/06
JAEA has developed plutonium and uranium mixed oxide (MOX) containing 20-32%Pu content as a fuel of the fast breeder reactor. During irradiation, large temperature gradient in radial direction of a fuel pellet causes redistribution of Pu and U, and the Pu content increases to about 43% at the pellet center. The maximum temperature of the fuel pellet during irradiation is limited within the design criterion to prevent fuel melting. So, it is important to evaluate melting points of MOX containing 43%Pu. In this work, it is confirmed that the MOX with 43%Pu content is not melted by heat treatment just below the melting point which was determined by thermal arrest technique using Re inner capsule. The MOX specimen with 43%Pu content was heated at 2978K for 40s using Re inner capsule. Optical micrograph and XRD results show the specimen was heated at the temperature less than solidus temperature. So it was confirmed that (PuAmU)O was solid phase at 2978K20K.
Kitamura, Akihiro; Nakamichi, Shinya; Okada, Takashi
Proceedings of 15th International Conference on Nuclear Engineering (ICONE-15) (CD-ROM), 7 Pages, 2007/04
Data on glovebox dismantling activities in the Glovebox Dismantling Facility (GDF) were analyzed to identify the work structure and the time consumed for each activity. As a result, we were able to categorize dismantling activities regarding time estimation point of view. The activities those of which variations are around 30% or less, were defined as "predictable activities", and activities those of which total time is small compare to the whole dismantling work were defined as "suppressible activities", and other as "unpredictable activities". In terms of these definition the time interval for unit activity were evaluated and found that almost all of the work can be predicted within 30% uncertainly.
Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Nakamichi, Shinya; Kashimura, Motoaki; Abe, Tomoyuki; Uno, Hiroki*; Ogasawara, Masahiro*; Tamura, Tetsuya*; Sugata, Hiromasa*; et al.
JAEA-Technology 2006-049, 32 Pages, 2006/10
JAEA-Technology-2006-049.pdf:19.46MBJAEA-Technology-2006-049(errata).pdf:0.32MB
Japan Atomic Energy Agency has developed a fast breeder reactor(FBR), and plutonium and uranium mixed oxide (MOX) having low density and 20-30%Pu content has used as a fuel of the FBR, Monju. In plutonium, Americium has been accumulated during long-term storage, and Am content will be increasing up to 2-3% in the MOX. It is essential to evaluate the influence of Am content on physical properties of MOX on the development of FBR in the future. In this study melting points and thermal conductivities which are important data on the fuel design were measured systematically in wide range of composition, and the effects of Am accumulated were evaluated. The solidus temperatures of MOX were measured as a function of Pu content, oxygen to metal ratio(O/M) and Am content using thermal arrest technique. The sample was sealed in a tungsten capsule in vacuum for measuring solidus temperature. In the measurements of MOX with Pu content of more than 30%, a rhenium inner capsule was used to prevent the reaction between MOX and tungsten. In the results, it was confirmed that the melting points of MOX decrease with as an increase of Pu content and increase slightly with a decrease of O/M ratio. The effect of Am content on the fuel design was negligible small in the range of Am content up to 3%. Thermal conductivities of MOX were evaluated from thermal diffusivity measured by laser flash method and heat capacity calculated by Nuemann- Kopp's law. The thermal conductivity of MOX decreased slightly in the temperature of less than 1173K with increasing Am content. The effect of Am accumulated in long-term storage fuel was evaluated from melting points and thermal conductivities measured in this study. It is concluded that the increase of Am in the fuel barely affect the fuel design in the range of less than 3%Am content.
Nakamichi, Shinya; Kitamura, Akihiro; Yanagawa, Chihiro; Nakai, Koji; Okada, Takashi; Asazuma, Shinichiro; Kashiro, Kashio
no journal, ,
no abstracts in English
Nakamichi, Shinya; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki; Sugata, Hiromasa*; Shibata, Katsuya*; Uno, Hiroki*; Tamura, Tetsuya*
no journal, ,
no abstracts in English
