Plutonium extraction from irradiated nuclear fuel

The solvent extraction process that uses TBP solutions to recover plutonium and uranium from irradiated nuclear fuels is called Purex (plutonium uranium extraction). The Purex process provides recovery of more than 99% of both uranium and plutonium with excellent decontamination of both elements from fission products. The Purex process is used worldwide to reprocess spent reactor fuel. During the last several decades, many variations of the Purex process have been developed and demonstrated on a plant scale. 

Since publication of this work, Japanese researchers have undertaken an effort to demonstrate the feasibility of direct dissolution of U02 from spent nuclear fuels by the TBP-HN03 complex in SC-C02.49 Ultimately, the project is directed at the extraction of both uranium and plutonium from mixed oxide fuels and from irradiated nuclear fuel. Ideally, soluble uranyl and plutonium nitrate complexes will form and dissolve in the C02 phase, leaving the FPs as unwanted solids. As in the conventional... 

Naylor, A. and Wilson, P. D. 1983. Recovery of uranium and plutonium from irradiated nuclear fuel. In Handbook of solvent extraction, eds. T. C. Lo, M. H. I. Baird, and C. Hanson, 768-798. New York John WUey Sons. 

The effect of irradiation on the extractability of sulfoxides towards plutonium, uranium and some fission products were studied by Subramanian and coworkers . They studied mainly the effect of irradiation on dihexyl sulfoxide (DHSO) and found that irradiation did not change the distribution coefficient for Ru, Eu and Ce but increases the distribution coefficient for Zr and Pu. When comparing DHSO and tributyl phosphate (TBP), the usual solvent for the recovery and purification of plutonium and uranium from spent nuclear fuels, the effect of irradiation to deteriorate the extraction capability is much larger in TBP. Lan and coworkers studied diphenyl sulfoxides as protectors for the gamma radiolysis of TBP. It was found that diphenyl sulfoxide can accept energy from two different kinds of excited TBP and thus inhibits the decomposition of the latter. 

Applicability of SFE to nuclear fuel reprocessing has been proposed by Smart and Wai et al. (17, 18). We have developed a new process which employs a high pressure mixture of TBP-HNO3-H2O-CO2 as is described in this chapter and this approach has indicated a very efEcient extraction of uranium from UO2. Now, the nuclear industries have paid attention to the applications of SFE to future processes. In Japan, we have started a four-year project with nuclear plant construction companies to demonstrate uranium and plutonium extraction from a mixed oxide fuel using the high pressure mixture. On the other hand, uranium and plutonium will be extracted from the irradiated nuclear fuel with TBP(HN03)i.s(H20)o.6 in the same project. 

The ability of diethyl ether to extract uranyl nitrate from aqueous solution has been known for a hundred years and was the method chosen by the Manhattan Project to purify the uranium used in the first nuclear chain reactors. This solvent has numerous disadvantages. It is very volatile, very flammable, and toxic, and it requires addition of sodium, aluminum, or calcium nitrate to the aqueous phase to enhance extractions. When solvent extraction was first applied to recovery of uranium and plutonium from irradiated fuel, other oxygenated solvents less volatile than diethyl ether that were first used were methyl isobutyl ketone, dibutyl... 

The reprocessing involves separating the fission products from the actinides, and then separating the plntoninm from the uranium. The best known procedure of this type is the PUREX (Plutonium, URanium Extraction) process that is used for recovery of uranium and plutonium from irradiated fuel (see details in Chapter 2). The separated plutonium can be used for the production of nuclear weapons or converted into the oxide form, mixed with nraninm oxide and can be used as MOX nuclear fuel. 

After the fuel is irradiated in the reactor, there are two divergent paths proposed to handle the used fuel. The first is to package it and dispose of it in a geologic repository as high-level nuclear waste. The alternative is to reprocess it to extract the remaining uranium and also recover the plutonium that was produced during irradiation. The waste from this process is the fission products. As seen in Chapter 14, there are several processes that can be used to preserve the useful uranium and plutonium. 

In principle, two fundamentally different methods can be applied to solve this task. The first one is determination of the residual concentrations of the fissile nuclides after irradiation and calculation of the burnup from the difference between final and initial values. For this purpose, the uranium and plutonium fraction has to be separated from the fission and activation products and from each other (e. g. by extraction chromatography) subsequently, the concentrations of the individual isotopes, in particular of the fissile isotopes, are analyzed by mass spectrometry. Well-established analytical techniques for performing such analyses are available, so that only small error margins are to be expected in the determination of the concentrations of the isotopes under consideration. However, there are two problems that can potentially cause systematic errors. The first one is the well-known question of the accuracy of results which have been obtained as a difference between two numbers, which limits the accuracy at lower burnup values in particular. The second problem is that the fissile nuclides are not only consumed by nuclear fission but by neutron capture as well in order to avoid systematic errors here, the capture-to-fission ratio valid for the particular irradiation conditions has to be taken into account in the calculation of depletion during irradiation. If one recalls the complicated buildup and decay mechanisms of actinide nuclides during reactor irradiation (see Fig. 3.5.), it is obvious that such correction requires complex calculations. On the other hand, the direct determination of the residual concentration of fissile nuclides is not influenced by errors due to inaccuracies in the fission yields of fission products to be measured nor by migration-induced inho-mogenities in the fuel. 

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