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Dissolution of Zircaloy-2-Clad UO2 Commercial Reactor Fuel

G. F. Kessinger, M. C. Thompson

Nuclear Technology / Volume 169 / Number 3 / Pages 263-270

March 2010

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The primary goal of this investigation was to evaluate the effectiveness of the chop-leach process, with nitric acid solvent, to produce a nominally 300 g/l [U] and 1 M [H+] product solution. The results of this study show that this processing technique is appropriate for applications in which a low free acid and moderately high U content are desired. The 7.75 l of product solution, which was >450 g/l in U, was successfully diluted to produce [approximately]13 l of solvent extraction feed that was 302 g/l in U with a [H+] in the range 0.8 to 1.2 M.

A secondary goal was to test the effectiveness of this treatment for the removal of actinides from Zircaloy cladding to produce a low-level radioactive waste (LLW) cladding product. Analysis of the cladding shows that actinides are present in the cladding at a concentration of [approximately]5000 Ci/g, which is about 50 times greater than the acceptable transuranium element limit in LLW.

It appears that the concentration of nitric acid used for this dissolution study (initial concentration 4 M, with 10 M added as the dissolution proceeded) was inadequate to completely digest the UO2 present in the spent fuel. The mass of insoluble material collected from the initial treatments with nitric acid, 340 g, was much higher than expected, and analysis of this insoluble residue showed that it contained at least 200 g U.

 
 
 
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