The EURO-GANEX Process: Current Status of Flowsheet Development and Process Safety Studies

A new hydrometallurgical grouped actinide extraction process has been developed to separate the transuranic actinide ions from dissolved spent fuel solution (after an initial uranium extraction cycle). This “EURO-GANEX” process is aimed towards the homogeneous recycling of plutonium and minor actinides in a future closed fuel cycle. The separation process is based on the co-extraction of actinides and lanthanides from aqueous nitric acid into an organic phase followed by selective co-stripping of actinides. A suitable organic phase has been formulated and distribution ratios determined for lanthanides, actinides and some problematic fission products under extraction and stripping conditions. The process flowsheet has been proven on surrogate feed solutions as well as with spent fast reactor fuel; excellent recoveries of the actinides and good decontamination factors from the lanthanides and other fission products were obtained. A variation on the EURO-GANEX flowsheet (the “TRU-SANEX” process) has now been designed to produce separate Pu+Np and Am+Cm products for heterogeneous recycling. Progress on underpinning process chemistry and safety studies as well as flowsheet tests are summarized.     

Sequential separation of actinides and lanthanides by extraction chromatography using a CMPO-TBP/XAD7 column

CMPO/TBP sorbed on Amberlite XAD7 resin was used for the separation of actinides and lanthanides from nitric acid solutions by extraction chromatography. The distribution ratios of actinides and lanthanide fission products (Ce, Eu) as a function of acid concentration and some complexing agents were determined. In strong HNO₃ medium (>1 mol/l) the tri-, tetra- and hexavalent actinides as well as the lanthanides have shown great affinity for the CMPO/TBP/XAD7 sorbent. The same behavior was found in HCl medium except for trivalent actinides and lanthanides which show lower distribution values in the same acid range. The effect of some complexing agents as DTPA and ammonium oxalate were also investigated. In DTPA only hexavalent actinides showed higher distribution value. On the basis of these differences, an alternative procedure for actinide-lanthanide separation and actinides from each other is proposed.     

Separation of actinides from Low Level Liquid Wastes (LLLW) by extraction chromatography using novel DMDOHEMA and TODGA impregnated resins

The uptake of several actinides [U(VI), Th(IV), Am(III), Cm(III)] and fission products was investigated from nitric acid solutions by two novel extraction chromatographic sorbents containing 2-(2-hexyloxy-ethyl)-N,N'-dimethyl-N,N'-dioctyl-malonamide (DMDOHEMA) and N,N,N',N'-tetraoctyl-3-oxapentane-1,5-diamide (TODGA), respectively. The kinetics of the uptake of actinides was studied. The sorption of metal ions fromz simulated Low Level Liquid Waste (LLLW) solutions was evaluated. The results of these experiments revealed that the actinides and lanthanides could be separated from the bulk of other fission products in simulated LLLW solutions on both sorbents.This revised version was published online in November 2005 with corrections to the Cover Date.     

A Bulk Resin Bead Procedure To Obtain Uranium And Plutonium From Radioactive Solutions For Mass Spectrometric Analysis

Abstract

A method is described for extracting representative uranium and plutonium samples from highly radioactive solutions for isotopic mass spectrometric analysis. Anion resin beads in the nitrate form are used to effect separation from fission products and other actinides. Conditions required to achieve separation are proper adjustment of the uranium and nitric acid concentrations. Once uranium and plutonium are adsorbed, each bead serves as a sample for mass spectrometric analysis, with plutonium and uranium being run sequentially from the same bead. Quantitative determination of the two elements is effected through the technique of isotopic dilution.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

Evaluation of tertiary pyridine resin for the separation of lanthanides by simulated moving‐bed chromatography

Lanthanide separation by simulated moving‐bed chromatography was studied as a model system for separating lanthanide fission products and minor actinides from used nuclear fuels. The simulated moving‐bed system was modeled for a tertiary pyridine anion‐exchange resin supported on silica particles as the stationary phase and a mixture of methanol and 1M nitric acid as the mobile phase. Pulse injection tests using a single packed column were used to obtain chromatographic parameters for mathematical modeling of the simulated moving‐bed system. Higher concentrations of methanol improved the separation, but the chromatograms showed evidence of nonlinearity of the isotherms. The mathematical model of the simulated moving‐bed process predicted a production rate of purified samarium and neodymium at 118 g solute/L resin/day and a purity of 99.5%. The optimal methanol ratio for the production rate for various product purities was determined from the model. The excellent separation of Nd and Sm suggests that the simulated moving‐bed system could be applied to the separation of minor actinides such as americium and curium.     

Final purification and concentration of americium/curium separated from high-level reprocessing waste

A cation-exchange cycle has been developed for the recovery and concentration of the Am/Cm product from a DTPA/lactic acid solution used in an extraction process for the isolation and separation of the actinides from lanthanide fission products. The optimum pH region for the sorption of Am³⁺ from 0.05M DTPA/1M lactic acid solutions by strongly acidic cation-exchange resins is pH 0.9–1.0. Maximum usable capacities, heights of the exchange zone, and concentration factors for different resins, cross-linkages and temperatures have been determined. Decontamination factors are given for some fission products, as well as U, Np and Pu.     

New conceptual process for the extraction of actinides based on oxapentane-diamides

Several diamide derivates were synthesized in our laboratory. The extraction of actinides and some fission products by these compounds were studied. N,N,N’,N’-tetra-(2-ethylhexyl)-3-oxa pentanediamide [TEHOPDA) was proven to be a suitable extractant for the removal of actinides from nitric acid solution. The actinides can be stripped from the loaded solvent by the dilute nitric acid. TEHOPDA showed a high loading capacity to actinides and lanthanides with a mixture of n-octanol and kerosene as the diluent. Considering the effective-extraction and easy-stripping of actinides, 0.25 mol/l TEHOPDA — 30% n-octanol + 70% kerosene was selected as the solvent. A cascade extraction experiment was carried out with the simulated dissolver solution of spend fuel as feed. 99.99% U and 99.999% Am, Pu, and Np were extracted in a 4-stage test. Based on the experimental results, a conceptual reprocessing process was proposed.     

Separation des actinides dans une solution renfermant des elements naturels et des produits de fission

A diagram for the separation of actinides is proposed. The task was to separate actinides from a soil sample, contaminated by them and fission products. Separations are performed by extraction chromatography and selective stripping followed by ion-exchange purification on small columns. The obtained actinides are free from foreign elements. It is possible to prepare electrodeposited sources for radiometric measurements, alpha-and gamma-countings or deposits for mass-spectrometric measurements.      

Optimizing composition of TODGA/SiO2-P adsorbent for extraction chromatography process

As a part of developing extraction chromatography technology for minor actinides (MA(III); Am and Cm) recovery from spent fast reactor fuels, improvement on the TODGA/SiO2-P adsorbent to enhance its desorption efficiency was carried out. Batchwise adsorption/elution experiments showed that optimizations in amount of the extractant impregnated in the support of the SiO2-P which is the porous silica coated with polymer and degree of the cross linkage of polymer succeeded in finding the optimum values. Inactive column separation experiments with the simulated high level liquid waste and the optimized adsorbent revealed that decontamination factors of fission products can also be improved as well as the recovery yields.     

Separation of gram quantities of uranium from fission products by extraction chromatography

The separation of gram quantities of uranium from fission products has been investigated by extraction chromatography. The separation which is based on the difference in distribution coefficients between uranium and the fission products on a tributyl phosphate (TBP) resin in nitric acid medium, was carried out by means of high acidity feed and stepwise elution on a TBP chromatography column. The results show that this technique is capable to separate 5 g of uranium from a large quantity of fission products. The recovery of uranium is more than 99%. The decontamination factors of g- and b-activities were 2.1^.10³ and 2.3^.10³, respectively.     

Reversed-phase partition chromatographic separation of minor actinides with bis(2-ethylhexyl) sulfoxide from acidic nitrate PUREX waste solutions

The uptake behavior of U(VI), Pu(IV), Am(III) and a few long-lived fission products from nitric acid media by bis(2-ethylhexyl) sulfoxide (BESO) adsorbed on Chromosorb has been studied U(VI), Pu(IV) and Zr(IV) are taken up appreciably as compared to trivalent actinides/lanthanides including some coexisting fission product contaminants which are weakly sorbed on the column. Chromosorb could be loaded with (1.12±0.03) g of BESO per g of the support. Maximum sorption is observed around 4–5 mol·dm^–3 HNO₃ for both U(VI) and Pu(IV), which are sorbed as their disolvates. The elution of (U(VI) and Pu(IV) from the metal loaded sorbent has also been optimized. Desorption of U(VI) is easily accomplished with dilute nitric acid (ca. 0.01 mol·dm^–3)while Pu(IV) is reductively stripped with 0.1 mol·dm^–3 NH₂OH·HCl. Effective sequential separation of U(VI), Pu(IV) and Am(III) from their several admixtures could be readily achieved from real medium and low level active acidic process raffinates.     

Carbon nanotubes: Potential uses in radionuclide concentration

The review of literature data related to the preparation, properties, and application of carbon nanotubes for sorption recovery of elements is given. Experimental data on the application of Taunit carbon nanofor radionuclide preconcentration from different solutions, as well as of Taunit-based solid-phase extractants for recovery of actinides and rare-earth elements from nitric acid solutions are presented.     

Effect of gamma-radiation on the extraction of europium and americium by benzyldimethyldodecylammonium nitrate

Previously it was found that in the extraction separation on lanthanides and americium from acidic nitrate solutions of nuclear fission products, benzyldimethyldodecylammonium nitrate gives high values of separation coefficients. The change in the extraction capacity of this agent and its solutions in benzene in the extraction of Eu(III) and Am(III) was investigated as a function of the adsorbed dose of ionizing radiation. The slight reduction in the extraction of both metals is caused mainly by the radiolysis products of nitric acid in the organic phase that enter into secondary reactions with both the solvent and the extractant. Comparison of the radiation stability of benzyldimethyldodecylammonium nitrate systems with tertiary amines show that the changes in distribution coefficients in the range of investigated absorbed doses are significantly lower in the former case. The investigated system may be characterized as radiation stable up to about 100 kGy even in the presence of nitric acid.     

Radioanalytical determination of actinides and fission products in Belarus soils

Alpha emitting actinides such as plutonium, americium or curium were measured by alpha-spectrometry after radiochemical separation. The short range of alpha-particles within matter requires, after a pre-concentration process, a succession of isolation and purification steps based on the valence states modification of the researched elements. For counting, actinides were electrodeposited in view to obtain the mass-less source necessary to avoid self-absorption of the emitted radiations. Activity concentrations of gamma-emitting fission products were calculated after measurement with high purity germanium detectors (HPGe). These different methods were used to analyse soils sampled in the Republic of Belarus, not far from the Chernobyl nuclear plant.     

Quantitative determination of tri-n-butyl phosphate by fluorescent X-ray spectrometry through inorganic association with an actinide

Tri-n-butyl phosphate (TBP) continues to be the most widely used solvent in nuclear fuel extraction, refining and reprocessing units for the extraction of actinides and their separation from fission products. An X-ray fluorescence spectrometric method (XRFS) for the determination of TBP content with an X-ray detectable element is presented. The method involves formation of an ion association complex of uranium with TBP-kerosene mixture in 3M nitric acid. The analytes uranium and bromine used as internal ratio elements in organic extract are excited by a primary X-ray beam from a rhodium tube. The solvent concentration is determined from the ratioed characteristic intensities of uranium and bromine. The procedure permits the determination of organic solvent in the range 0.5 to 5.0% with a relative standard deviation of 0.1%.     

Recovery of actinides from acidic waste solutions generated in research facilities using adsorption and extraction techniques

Journal of Radioanalytical and Nuclear Chemistry - In the present study various separation methods were investigated for the recovery of actinides from aqueous waste solutions. Extraction behavior...     

Extraction-chromatographic separation of uranium from long-lived fission products using tributyl phosphate

Extraction-chromatographic separation of uranium from fission products was performed using undiluted tributyl phosphate sorbed on Chromosorb W as a stationary phase, and nitric acid (1: 3) as a mobile phase. Most of the fission products that contributed greatly to the radiation level of the sample passed through the column; this effected considerable decontamination. Uranium retained on the column was quantitatively recovered by elution with water.     

A pulse radiolysis investigation of the reactions of tributyl phosphate with the radical products of aqueous nitric acid irradiation

Tributyl phosphate (TBP) is the most common organic compound used in liquid-liquid separations for the recovery of uranium, neptunium, and plutonium from acidic nuclear fuel dissolutions. The goal of these processes is to extract the actinides while leaving fission products in the acidic, aqueous phase. However, the radiolytic degradation of TBP has been shown to reduce separation factors of the actinides from fission products and to impede the back-extraction of the actinides during stripping. As most previous investigations of the radiation chemistry of TBP have focused on steady state radiolysis and stable product identification, with dibutylphosphoric acid (HDBP) invariably being the major product, here we have determined room temperature rate constants for the reactions of TBP and HDBP with the hydroxyl radical [(5.00 +/- 0.05) x 10(9), (4.40 +/- 0.13) x 10(9) M(-1) s(-1)], hydrogen atom [(1.8 +/-0.2) x 10(8), (1.1 +/- 0.1) x 10(8) M(-1) s(-1)], nitrate radical [(4.3 +/- 0.7) x 10(6), (2.9 +/- 0.2) x 10(6) M(-1) s(-1)], and nitrite radical (<2 x 10 (5), <2 x 10(5) M(-1) s(-1)), respectively. These data are used to discuss the mechanism of TBP radical-induced degradation.     

Potentialities of fluoride-based salts for specific nuclear reprocessing: Overview of the R&D program at CEA

Initially studied in the frame of the first French act on radioactive waste management (December 1991), the pyrotechnology is currently assessed by the Nuclear Energy Direction of the Commissariat à l’Energie Atomique (CEA) within the succeeding act (June 2006) as a potential alternative to hydrometallurgy for the reprocessing of targets or dedicated fuels (coming from accelerator-driven systems or ADS) considered for the minor actinides transmutation.The R&D program is mainly focused on the evaluation of the fluoride melts as interesting media for operating separation between the actinides and the fission products. Two separation techniques are currently evaluated; the first one uses the liquid-liquid extraction technique between molten fluoride and liquid metal at high temperature, the second one is based on an electrolytic separation in a molten fluoride melt. Both are promising in terms of separation efficiency. This paper gives an overview of the current studies and presents the last main experimental results.