After a brief description of the PUREX process as a starting point, this paper aims to point out some of its most recent developments, always from the viewpoint of the chemistry involved in the solvent extraction methods. In addition to PUREX, other well-known technological approaches devoted to the reprocessing of irradiated uranium and thorium nuclear fuels and general radioactive wastes — TRUEX, DIAMEX, TRPO, DIDPA and UNEX — are also presented and discussed.

Recent advances on the chemistry of solvent extraction applied to the reprocessing of spent nuclear fuels and radioactive wastes

The partitioning of the long‐lived α‐emitters and the high‐yield fission products from dissolved used nuclear fuel is a key component of processes envisioned for the safe recycling of used nuclear fuel and the disposition of high‐level waste. These future processes will likely be based on aqueous solvent‐extraction technologies for light‐water reactor fuel and consist of four main components for the sequential separation of uranium, fission products, group trivalent actinides, and lanthanides, and then trivalent actinides from lanthanides. Since the solvent systems will be in contact with highly radioactive solutions, they must be robust toward radiolytic degradation in an irradiated mixed organic/aqueous acidic environment, with the formation of only benign degradation products. Therefore, an understanding of their radiation chemistry is important to the design of a practical system. In the first paper in this series, we reviewed the radiation chemistry of irradiated aqueous nitric acid and the tributyl ...

Review Article: The Effects of Radiation Chemistry on Solvent Extraction 3: A Review of Actinide and Lanthanide Extraction

The TRUEX (TRansUranic EXtraction) solvent extraction process was developed at Argonne National Laboratory (ANL) for the Department of Energy. A TRUEX demonstration completed at ANL involved the processing of analytical and experimental waste generated there and at the New Brunswick Laboratory. A 20-stage centrifugal contactor was used to recover plutonium, americium, and uranium from the waste. Approximately 84 g of plutonium, 18 g of uranium, and 0.2 g of americium were recovered from about 118 L of solution during four process runs. Alpha decontamination factors as high as 65,000 were attained, which was especially important because it allowed the disposal of the process raffinate as a lowlevel waste. The recovered plutonium and uranium were converted to oxide; the recovered americium solution was concentrated by evaporation to approximately 100 mL. The flowsheet and operational procedures were modified to overcome process difficulties. These difficulties included the presence of complexants i...

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TRUEX Processing of Plutonium Analytical Solutions at Argonne National Laboratory

The TRUEX (TRansUranic EXtraction) solvent extraction process was developed at Argonne National Laboratory (ANL) for the Department of Energy. A TRUEX demonstration completed at ANL involved the processing of analytical and experimental waste generated there and at the New Brunswick Laboratory. A 20-stage centrifugal contactor was used to recover plutonium, americium, and uranium from the waste. Approximately 84 g of plutonium, 18 g of uranium, and 0.2 g of americium were recovered from about 118 liters of solution during four process runs. Alpha decontamination factors as high as 65,000 were attained, which was especially important because it allowed the disposal of the process raffinate as a low-level waste. The recovered plutonium and uranium were converted to oxide; the recovered americium solution was concentrated by evaporation to approximately 100 ml. The flowsheet and operational procedures were modified to overcome process difficulties. These difficulties included the presence of complexants in the feed, solvent degradation, plutonium precipitation, and inadequate decontamination factors during startup. This paper will discuss details of the experimental effort.

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TRUEX processing of plutonium analytical solutions at Argonne National Laboratory

An extractant system, 0.75 MDHDECMP and 1 M TBP in an aliphatic hydrocarbon diluent, was evaluated for selectively removing transuranic elements (TRUs) from acidic waste streams. This evaluation involved six parts: 1) Investigating the tendency to form third phases as a function of HNO3 concentration. 2) Measuring the distribution coefficients for the radioactive waste components UO2 2+, Am3+ and Pu4+ as a function of aqueous nitrate concentration. 3) Measuring the distribution coefficients for the important, nonradioactive waste components HNO3, Fe3+, Zr4+, and Bi3+ as a function of aqueous nitrate concentration. 4) Qualitatively determining the impact of potential feed additives such as fluoride ion and oxalate ion upon these distribution values. 5) Qualitatively evaluating the tendency to form interfacial crud upon contact with acidified solutions containing high metal (Zr4+) loadings. 6) Investigating extractant pretreatment to remove trace organophosphorus impurities which affect metal distr...

Extraction of transuran1c elements with a dihexyl-n.n-diethylcarbamoylmethyl phosphonate/tributyl phosphate system - initial investigations

Solid sorbents, alumina, silica gel, and Amberlyst A-26 have been tested for the cleanup of degraded TRUEX-NPH solvent A sodium carbonate scrub alone does not completely remove acidic degradation products from highly degraded solvent and cannot restore the stripping performance of the solvent By following the carbonate scrub with either neutral alumina or Amberlyst A-26 anion exchange resin, the performance of the TRUEX-NPH is substantially restored The degraded TRUEX-NPH was characterized before and after treatment by supercritical fluid chromatography Its performance was evaluated by americium distribution ratios, phase-separation times, and lauric acid distribution coefficients 17 refs, 2 figs, 5 tabs

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Lucinda Reichley-yinger

Papers

Truex Process Solvent Cleanup with Solid Sorbents