Residual Salt Separation from Simulated Spent Nuclear Fuel Reduced in a LiCl–Li2O Salt

JOURNAL OF CHEMICAL ENGINEERING OF JAPAN

Jin-Mok Hur¹  Chung-Seok Seo¹  Sun-Seok Hong¹ 

[1] Korea Atomic Energy Research Institute

关键词: Separation;    Spent Nuclear Fuel;    Electrochemical Reduction;    Molten Salt;    Vaporization;   

DOI  :  10.1252/jcej.39.1291

来源: Maruzen Company Ltd

【 摘 要 】

References(8)Cited-By(1)The electrochemical reduction of spent nuclear fuel in LiCl–Li2O molten salt for the conditioning of spent nuclear fuel requires the separation of the residual salts from a reduced metal product after the reduction process. Considering the behavior of spent nuclear fuel during the electrochemical reduction process, a surrogate material matrix was constructed and inactive tests on a salt separation were carried out to produce the data required for active tests. Fresh uranium metal prepared from the electrochemical reduction of U3O8 powder was used as the surrogates of the spent nuclear fuel components which might be metallized by the electrochemical reduction process. LiCl, Li2O, Y2O3 and SrCl2 were selected as the components of the residual salts. Interactions between the salts and their influence on the separation of the residual salts were analyzed by differential scanning calorimetry (DSC) and thermogravimetry (TG). Eutectic melting of LiCl–Li2O and LiCl–SrCl2 led to a melting point which was lower than that of the LiCl molten salt was observed. Residual salts were separated by a vaporization method. Co-vaporization of LiCl–Li2O and LiCl–SrCl2 was achieved below the temperatures which could make the uranium metal oxidation by Li2O possible. The salt vaporization rates at 950°C were measured as follows: LiCl–8 wt% Li2O > LiCl > LiCl–8 wt% SrCl2 > SrCl2.

【 授权许可】

Unknown