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

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

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.

Language	:	English
Journal	:	Transactions of the American Nuclear Society
Volume	:	97
Number	:	1
Pages	:	p.618 - 619
Publication Year/Month	:	2007/11
Meeting title	:	ANS/ENS International Meeting and Nuclear Technology Expo 2007
Held date	:	2007/11
Location (city)	:	Washington, D.C.
Location (country)	:	U.S.A.
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Keywords	:	no keyword
Research Facility	:	プルトニウム燃料製造施設
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Registration No. : BB20070706
JAEA Abstracts No. : 36000089
Paper Submission No. : 3415

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