- P-ISSN 1225-0163
- E-ISSN 2288-8985
- P-ISSN1225-0163
- E-ISSN2288-8985
- SCOPUS, ESCI, KCI
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A Study on the Separation of Neodymium from the Simulated Solution of <TEX>$U_3Si/Al$</TEX> Spent Nuclear Fuel
Analytical Science and Technology / Analytical Science and Technology, (P)1225-0163; (E)2288-8985
2000, v.13 no.5, pp.584-591
Choi,
K. S., Kim,
J. S., Han,
S. H., Park,
S. D., Park,
Y. J., Joe,
K. S., &
Kim,
W.
H.
(2000). A Study on the Separation of Neodymium from the Simulated Solution of <TEX>$U_3Si/Al$</TEX> Spent Nuclear Fuel. Analytical Science and Technology, 13(5), 584-591.
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Abstract
The separation of Nd from the simulated <TEX>$U_3Si/Al$</TEX> spent fuel solution with sequential two-step anion exchange separation has been studied. To prepare the simulated <TEX>$U_3Si/Al$</TEX> spent nuclear fuel, unirradiated <TEX>$U_3Si/Al$</TEX> whose composition consists of small <TEX>$U_3Si$</TEX> particle dispersed in an Al matrix with Al cladding was dissolved with a mixture of 4 M HCl and 10 M <TEX>$HNO_3$</TEX> and 8 or 15 fission product elements were added to the dissolved solution. The trace amount of silica in the solutions was removed by evaporating to dryness with HF and the U was adsorbed on the first anion exchange resin. Neodymium can be purely isolated from the fission product elements with a methanol-nitric acid eluent using the second anion exchange resin. A large excess of Al didn't influence on the elution velocity of Nd, but reduced the eluted contents of Nd, Al, Eu, Gd, Sm and Sr, A large amount of Al was removed first from the column with 3 mL of loading solution (0.8 M <TEX>$HNO_3$</TEX>/99.8% MeOH) before Nd elution by the eluent [0.04 M <TEX>$HNO_3$</TEX>-99.8% MeOH(1:9)]. The recovery of Nd was more than 94%, regardless of Al contents. Taking the 9 to 13 mL fraction of eluate was effective to purely isolate Nd.
- keywords
- neodymium separation, <TEX>$U_3Si/Al$</TEX> fuel, anion exchange chromatography
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