DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL

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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{sup +}] 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 over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 M. ... continued below

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Kessinger, G. & Thompson, M. August 7, 2009.

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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{sup +}] 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 over 450 g/L in U, was successfully diluted to produce about 13 L of solvent extraction feed that was 302 g/L in U with a [H{sup +}] in the range 0.8-1.2 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 about 5000 {eta}Ci/g, which is about 50 times greater than the acceptable transuranium element limit in low level radioactive waste. 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 UO{sub 2} 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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  • Journal Name: Nuclear Technology

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  • Report No.: SRNL-STI-2009-00496
  • Grant Number: DE-AC09-08SR22470
  • Office of Scientific & Technical Information Report Number: 962686
  • Archival Resource Key: ark:/67531/metadc930778

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  • August 7, 2009

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  • Nov. 13, 2016, 7:26 p.m.

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  • Dec. 12, 2016, 3:47 p.m.

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Kessinger, G. & Thompson, M. DISSOLUTION OF ZIRCALOY 2 CLAD UO2 COMMERCIAL REACTOR FUEL, article, August 7, 2009; South Carolina. (digital.library.unt.edu/ark:/67531/metadc930778/: accessed October 22, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.