Development of electrochemical reduction process for oxide nuclear fuel, 9; Solubility measurement of Pu $_{2}$ O $_{3}$ in molten LiCl-Li $_{2}$ O

Kato, Tetsuya*; Sakamura, Yoshiharu*; Iwai, Takashi; Arai, Yasuo

In order to apply the pyrometallurgical reprocessing to spent oxide fuels, reduction techniques of oxide in molten salt have been studied. In the previous reduction experiments, a small amount of Pu has been found to move out of the reduced fuel and dissolve into molten LiCl-Li $_{2}$ O. In this study, solubility of Pu in molten LiCl-Li $_{2}$ O was measured to obtain quantitative data for the Pu dissolution. The Pu solubility in the molten LiCl-Li $_{2}$ O containing Pu $_{2}$ O $_{3}$ precipitates increased with increasing Li $_{2}$ O concentration in the molten salt to be approximately 0.3wt%Pu at 8wt%Li $_{2}$ O and 923 K.

Language	:	Japanese
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Meeting title	:	日本原子力学会2007年秋の大会
Held date	:	2007/09
Location (city)	:	北九州
Location (country)	:	日本
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Keywords	:	no keyword
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Cooperating Institute	:	電力中央研究所

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Registration No. : BB20071162
JAEA Abstracts No. :
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Development of electrochemical reduction process for oxide nuclear fuel, 9; Solubility measurement of PuO in molten LiCl-LiO

Development of electrochemical reduction process for oxide nuclear fuel, 9; Solubility measurement of Pu $_{2}$ O $_{3}$ in molten LiCl-Li $_{2}$ O