Chromatographic Separation of Metals

Abstract

In nuclear reprocessing, PUREX, a solvent extraction process, has long been the separation method employed for the separation of the bulk components of irradiated nuclear fuel (namely uranium and plutonium) from the fission products and other minor actinides produced during the fuel use. The uranium and plutonium constitutes approximately 96 % by mass of the irradiated fuel and for this to be removed, requires large volumes of extractant and equipment with large surface area contactors and therefore floor space requirements.
The PUREX process has for nearly 60 years been the largely unchallenged separation technology for the reprocessing of irradiated fuel, for both nuclear weapon production and commercial nuclear power generation. The merits and ability of this process are unquestionable since it achieves the objectives of highly purified plutonium and uranium which both can be eventually recycled.
Although well proven and predictable, the PUREX process is not without its challenges: the generation of significant quantities of highly active aqueous liquid containing fission products (FPs) and minor actinides (MAs), and the degradation of the solvent phase reagents and non-specific nature of the extractant TriButylPhosphate (TBP) may have contributed to only a fraction of the total annual output of irradiated fuel being reprocessed.
Fission products are elements which are produced in a nuclear reactor and are the atomic fragments left after a large atomic nucleus (typically uranium-235) undergoes nuclear fission, splitting into two smaller nuclei, along with a few neutrons, the release of heat energy (kinetic energy of the nuclei), and gamma rays.
Minor actinides such as neptunium, americium, curium, berkelium, californium, einsteinium, and fermium are the actinide elements in irradiated nuclear fuel other than uranium and plutonium; they are minor as they represent a very small proportion of actinides in comparison to U and Pu.
This thesis explores the possibility of using a continuous chromatographic method to extract the lesser components of the irradiated fuel. One of the major problems with the use of chromatography as an industrial process is the expansion from the batch separations on the bench top to a continuous efficient process, capable of processing large volumes. This thesis, through existing concepts, will describe a proof of concept chromatographic separation of surrogates and isotopes of the components of irradiated fuels which can be readily scaled up to a continuous chromatographic separation.
The project is a radical departure from PUREX and will offer many advantages over PUREX. It is based on the separation of FPs and MAs from uranium and plutonium isotopes using continuous chromatographic separation.
This thesis assesses a number of commercial resins for their suitability for the proposed continuous chromatography reprocessing method. The experiments were all undertaken at elevated nitric acid concentrations and as such are describing interactions which are rarely required commercially and therefore seldom reported, with batch studies to assess separation factors between ions, uptake kinetics and isotherms over a range of nitric acid concentrations to more dynamic column breakthrough and eventually separations.
The research demonstrates that a separation can be achieved at an elevated HNO3 concentration on a commercially available ion exchange resin.