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Journal Articles

Development of science-based fuel technologies for Japan's Sodium-Cooled Fast Reactors

Kato, Masato; Hiroka, Shun; Ikusawa, Yoshihisa; Takeuchi, Kentaro; Akashi, Masatoshi; Maeda, Koji; Watanabe, Masashi; Komeno, Akira; Morimoto, Kyoichi

Proceedings of 19th Pacific Basin Nuclear Conference (PBNC 2014) (USB Flash Drive), 12 Pages, 2014/08

Uranium and plutonium mixed oxide (MOX) fuel has been developed for Japan sodium-cooled fast reactors. Science based fuel technologies have been developed to analyse behaviours of MOX pellets in the sintering process and irradiation conditions. The technologies can provide appropriate sintering conditions, irradiation behaviour analysis results and so on using mechanistic models which are derived based on theoretical equations to represent various properties.

Journal Articles

Effect of oxygen-to-metal ratio on properties of corium prepared from UO$$_{2}$$ and zircaloy-2

Hiroka, Shun; Kato, Masato; Morimoto, Kyoichi; Komeno, Akira; Uchida, Teppei; Akashi, Masatoshi

Journal of Nuclear Materials, 437(1-3), p.130 - 134, 2013/06

 Times Cited Count:4 Percentile:35.61(Materials Science, Multidisciplinary)

Journal Articles

Melting temperature and thermal conductivities of corium prepared from UO$$_{2}$$ and zircalloy-2

Kato, Masato; Uchida, Teppei; Hiroka, Shun; Akashi, Masatoshi; Komeno, Akira; Morimoto, Kyoichi

Materials Research Society Symposium Proceedings, Vol.1444, p.91 - 96, 2012/09

 Times Cited Count:1 Percentile:59.1

Journal Articles

Oxygen potentials of PuO$$_{2-x}$$

Komeno, Akira; Kato, Masato; Hiroka, Shun; Sunaoshi, Takeo*

Materials Research Society Symposium Proceedings, Vol.1444, p.85 - 89, 2012/09

 Times Cited Count:6 Percentile:95.04

Journal Articles

Phase separation behaviour of (U$$_{0.7}$$Pu$$_{0.3}$$)O$$_{2-x}$$ (1.92$$<$$x$$<$$2.00) based fuels containing actinides and/or lanthanides

Komeno, Akira; Kato, Masato; Uno, Hiroki*; Takeuchi, Kentaro; Morimoto, Kyoichi; Kashimura, Motoaki

IOP Conference Series; Materials Science and Engineering, 9, p.012016_1 - 012016_7, 2010/05

 Times Cited Count:8 Percentile:94.52

It is expected that the important data for design of fast reactor fuel can be provided by evaluating the relationship between fuel composition and phase separation with reported and new measurement data. According to evaluation with reported data and new measured data, a relationship between fuel composition and phase separation temperature of MOX fuel was indicated. Higher minor actinides-containing MOX had a lower phase separation temperature at O/M ratio region from 1.92 to 1.96.

Journal Articles

Experimental evaluation of Am-and Np-bearing mixed-oxide fuel properties

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.

Journal Articles

Self-radiation damage in plutonium and uranium mixed dioxide

Kato, Masato; Komeno, Akira; Uno, Hiroki*; Sugata, Hiromasa*; Nakae, Nobuo; Konashi, Kenji*; Kashimura, Motoaki

Journal of Nuclear Materials, 393(1), p.134 - 140, 2009/08

 Times Cited Count:33 Percentile:90.74(Materials Science, Multidisciplinary)

In plutonium compounds, the lattice parameter increases due to self-radiation damage by $$alpha$$-decay of plutonium isotopes. The lattice parameter change and its thermal recovery in plutonium and uranium mixed dioxide (MOX) were studied. The lattice parameter for samples of MOX powders and pellets that had been left in the air for up to 32 years was measured. The lattice parameter increased and was saturated at about 0.29%. The change in lattice parameter was formulated as a function of self-radiation dose. Three stages in the thermal recovery of the damage were observed in temperature ranges of below 673K, 673-1073K and above 1073K. The activation energies in each recovery stage were estimated to be 0.12 eV, 0.73 eV and 1.2 eV, respectively, and the corresponding mechanism for each stage was considered to be the recovery of the anion Frenkel defect, the cation Frenkel defect and a defect connected with helium, respectively.

Journal Articles

Evaluation of thermal conductivity of (U, Pu, Am)O$$_{2-x}$$

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki

Transactions of the American Nuclear Society, 97(1), p.618 - 619, 2007/11

Plutonium and uranium mixed oxide (MOX) fuel with high Pu-content has been developed as a fuel for fast reactors (FRs). Thermal conductivity of the oxide fuel is among the most important properties for design and performance analyses of fuel rods. Among recent reports, there have been none examining of thermal conductivity of MOX fuel containing Am except our studies. In this study, the thermal conductivities of MOX fuel with 30% Pu-content, as obtained by our group, were evaluated as functions of temperature, oxygen-to-metal (O/M) ratio and Am-content.

Journal Articles

Measurement of thermal conductivity of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ in high temperature region

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Ogasawara, Masahiro*; Sunaoshi, Takeo*

Transactions of the American Nuclear Society, 97(1), p.616 - 617, 2007/11

no abstracts in English

JAEA Reports

Evaluation of thermal physical properties for fast reactor fuels; Melting point and thermal conductivities

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:19.46MB
JAEA-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.

Oral presentation

Thermal physical properties of MOX Fuels, 3; Measurements of thermal diffusivity for (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{x}$$ (x=1.90$$sim$$2.00)

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

To examine the influences of O/M ratio on the thermal conductivity of MOX fuel, the thermal diffusivities of the MOX fuel with 30% Pu content were measured, and thermal conductivities were evaluated.

Oral presentation

Thermal physical properties of MOX fuels, 2; Thermal diffusivity measurement of (U, Pu, Am)O$$_{2.00}$$

Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Enthalpy measurement and heat capacity evaluation of (U, Pu, Am)O$$_{2}$$

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Thermal conductivity of (U$$_{0.68}$$Pu$$_{0.3}$$Am$$_{0.02}$$)O$$_{2-x}$$ at temperatures from 900 to 2200K

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Abe, Tomoyuki; Ogasawara, Masahiro*; Sunaoshi, Takeo*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Influence of self irradiation in MOX pellets stored over several decades

Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki; Sugata, Hiromasa*; Shibata, Katsuya*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Thermal conductivity of (U$$_{0.7-y}$$Pu$$_{0.3}$$Np$$_{y}$$)O$$_{2}$$

Morimoto, Kyoichi; Kato, Masato; Komeno, Akira; Kashimura, Motoaki; Ogasawara, Masahiro*

no journal, , 

no abstracts in English

Oral presentation

Thermal dependence of Density on Long-term stored MOX fuel; Variation of density and microstructure caused by heat treatment in MOX pellet stored for long term

Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Sugata, Hiromasa*; Shibata, Katsuya*; Uno, Hiroki*; Tamura, Tetsuya*

no journal, , 

no abstracts in English

Oral presentation

Property change and thermal recovery on plutonium and uranium mixed oxides by $$alpha$$ irradiation

Komeno, Akira; Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Sugata, Hiromasa*; Shibata, Katsuya*; Tamura, Tetsuya*; Uno, Hiroki*

no journal, , 

no abstracts in English

Oral presentation

Development of simplified pelletizing fuel production technology, 8; Physical properties of TRU-fuel produced

Segawa, Tomomi; Nakamichi, Shinya; Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Suzuki, Masahiro; Kihara, Yoshiyuki

no journal, , 

JAEA has been researching in the fundamental properties of MOX fuel including Minor Actinides(Am and Np). In this presentation, melting point, thermal conductivity, oxygen potential and lattice constant of (Am, Np, Pu, U)O$$_{2}$$ are reported.

Oral presentation

Thermal recovery behavior of thermal conductivity in MOX pellet stored for long-term

Komeno, Akira; Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Ogasawara, Masahiro*

no journal, , 

no abstracts in English

34 (Records 1-20 displayed on this page)