solid solutionsMorimoto, Kyoichi 
; Kato, Masato 
; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki
Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). As the storage time of Pu raw material between reprocessing and fabrication increases, americium content of the fabricated MOX fuel increases up to a few percent. In this work, the thermal conductivity of MOX fuel containing Am was investigated as a part of clarifying the effect of Am content on thermal physical properties. The pellets of (Am
Pu
U)O
, (Am
Pu U)O and (Am
Pu U)O were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. The thermal diffusivity measurement was carried out in the range of temperature from 900 K to 1700 K by the laser flash method, and thermal conductivity of these pellets was evaluated. The heat capacity for evaluating thermal conductivity was derived from heat capacity of UO, PuO and AmO by using the Kopp-Neumann rule.
Language |
: | English |
|---|---|---|
Journal |
: | |
Volume |
: | 452 |
Number |
: | 1 |
Pages |
: | p.54 - 60 |
Publication Year/Month |
: | 2008/03 |
Meeting title |
: | 12th International IUPAC Conference on High Temperature Materials Chemistry (HTMC-12) |
Held date |
: | 2006/09 |
Location (city) |
: | Vienna |
Location (country) |
: | Austria |
Paper URL |
: |
|
Keywords |
: | no keyword |
Research Facility |
: | |
Article of JAEA R&D Review |
: | Study on Physical Properties of Fast Reactor Fuel for Future Commenricialization; Study on Thermal Conductivities of Uranium-Plutonium Mixed Oxide Fuels[ |
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Percentile:77.85 Category:Chemistry, Physical |
Altmetrics |
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