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Oral presentation

Thermal conductivity of (U,Pu,Am)O$$_{2-x}$$ solid solutions

Morimoto, Kyoichi; Kato, Masato; Kashimura, Motoaki; Abe, Tomoyuki

no journal, , 

Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). In a current research, there is few research of the thermal conductivity evaluation of the MOX fuel with over 20% Pu content, and there is no research of thermal conductivity of the MOX fuel containing Am. In this work, to examine the influences of density, O/M ratio and Am content on the thermal conductivity of MOX fuel, the thermal diffusivities of the MOX fuel with 30% Pu content were measured, and the thermal conductivities of these MOX samples were evaluated.

Oral presentation

Measurements of melting points in the system PuO$$_{2}$$-UO$$_{2}$$

Kato, Masato; Morimoto, Kyoichi; Kashimura, Motoaki; Abe, Tomoyuki

no journal, , 

no abstracts in English

Oral presentation

Molecular dynamics simulations of MOX fuels containing minor actinides (MA: Np, Am)

Katayama, Masahito*; Adachi, Jun*; Kurosaki, Ken*; Osaka, Masahiko; Miwa, Shuhei; Tanaka, Kenya; Muta, Hiroaki*; Uno, Masayoshi*; Yamanaka, Shinsuke*

no journal, , 

no abstracts in English

Oral presentation

The FP (FP: Nd, Pd, Mo) effects on the thermal properties of the nitride fuel

Uno, Masayoshi*; Kurosaki, Ken*; Yamanaka, Shinsuke*; Minato, Kazuo

no journal, , 

no abstracts in English

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