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  • Author: Giuseppe Modolo x
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Abstract

The LUCA process was developed at Forschungszentrum Jülich for the selective separation of Am(III) from an acidic solution containing the trivalent actinides Am(III), Cm(III), and Cf(III) as well as lanthanides. A mixture of 0.4 mol/L bis(chlorophenyl)dithiophosphinic acid and 0.15 mol/L tris(2-ethylhexyl)phosphate dissolved in 20% isooctane/80% tert-butyl benzene was used as the extractant. The process was carried out in centrifugal contactors using an optimized flowsheet involving 7 stages for extraction, 9 stages for scrubbing and 8 stages for back-extraction. Very encouraging results were obtained. A high feed decontamination factor was obtained for Am(III) (>1000), and recovery in the product after stripping was higher than 99.8%. The Am(III) product was contaminated with 0.47% Cm(III). More than 99.9% Cf(III), Eu(III) and >99.5% Cm(III) inventories were directed to the raffinate and the contamination with Am(III) (<0.08%) was low. The experimental results were in good agreement with the predictions of a computer code.

Abstract

This work focuses on the reprocessability of metallic 92Mo and ceramic MgO, which is under investigation for (Pu,MA)-oxide (MA = minor actinide) fuel within a metallic 92Mo matrix (CERMET) and a ceramic MgO matrix (CERCER). Magnesium oxide and molybdenum reference samples have been fabricated by powder metallurgy. The dissolution of the matrices was studied as a function of HNO3 concentration (1-7 mol/L) and temperature (25-90°C). The rate of dissolution of magnesium oxide and metallic molybdenum increased with temperature. While the MgO rate was independent of the acid concentration (1-7 mol/L), the rate of dissolution of Mo increased with acid concentration. However, the dissolution of Mo at high temperatures and nitric acid concentrations was accompanied by precipitation of MoO3. The extraction of uranium, americium, and europium in the presence of macro amounts of Mo and Mg was studied by three different extraction agents: tri-n-butylphosphate (TBP), N,Nʹ-dimethyl-N,Nʹ-dioctylhexylethoxymalonamide (DMDOHEMA), and N,N,N’,N’- -tetraoctyldiglycolamide (TODGA). With TBP no extraction of Mo and Mg occurred. Both matrix materials are partly extracted by DMDOHEMA. Magnesium is not extracted by TODGA (D < 0.1), but a weak extraction of Mo is observed at low Mo concentration.

Abstract

The highly selective nitrogen donor ligand CyMe4BTBP for An(III) separation by solvent extraction was irradiated in a 60Co γ-source under varying conditions. Organic solutions of 10 mmol/L ligand in 1-octanol were contacted with different concentrations of nitric acid to observe the influence of an aqueous phase during irradiation. In subsequent liquid-liquid extraction experiments, distribution ratios of 241Am and 152Eu were determined. Distribution ratios decreased with increasing absorbed dose when irradiation was performed in the absence of nitric acid. With addition of nitric acid, initial distribution ratios remained constant over the whole examined dose range up to 300 kGy. For qualitative determination of radiolysis products, HPLC-MS measurements were performed. The protective effect of nitric acid was confirmed, since in samples irradiated with acid contact, no degradation products were observed, but only addition products of the 1-octanol molecule to the CyMe4BTBP molecule.

Abstract

Within the framework of the European collaborative project ACSEPT, a new SANEX partitioning process was developed at Forschungszentrum Jülich for the separation of the trivalent minor actinides americium, curium and californium from lanthanide fission products in spent nuclear fuels. The development is based on batch solvent extraction studies, single-centrifugal contactor tests and on flow-sheet design by computer code calculations. The used solvent is composed of 6,6´-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydrobenzo-[1,2-4]trizazin-3-yl)-[2,2´]-bipyridine (CyMe4BTBP) and N,N,N´,N´-tetraoctyldiglycolamide (TODGA) dissolved in n-octanol. A spiked continuous counter-current test was carried out in miniature centrifugal contactors with the aid of a 20-stage flow-sheet consisting of 12 extraction, 4 scrubbing and 4 stripping stages. A product fraction containing more than 99.9% of the trivalent actinides Am(III), Cm(III) and Cf(III) was obtained. High product/feed decontamination factors >1000 were achieved for these actinides. The trivalent lanthanides were directed to the raffinate of the process with the actinide (III) product stream being contaminated with less than 0.5 mass-% in the initial lanthanides.

Abstract

The novel hydrophilic back-extraction agent TS-BTPhen (3,3ʹ,3ʺ,3ʹʺ-[3-(1,10-phenanthroline-2,9-diyl)-1,2,4-triazine-5,5,6,6-tetrayl]tetrabenzenesulfonic acid) was tested for its selectivity towards Am(III) over Cm(III) and Eu(III) with a TODGA (N,N,Nʹ,Nʹ-tetraoctyldiglycolamide) based solvent. Batch experiments were carried out using TS-BTPhen dissolved in aqueous nitric acid solution with tracers of 152Eu, 241Am and 244Cm. A significant increase of the separation factor for Cm over Am from SFCm/Am = 1.6 up to SFCm/Am = 3.3 was observed compared to the use of a TODGA-nitric acid system alone. Furthermore, stripping was possible at high nitric acid concentrations (0.6-0.7 mol/L) resulting in a low sensitivity to acidity changes. The influence of the TS-BTPhen concentration was analyzed. A slope of -2 was expected taking into account literature stoichiometries of the lipophilic analogue CyMe4BTPhen. However, a slope of -1 was found. Batch stripping kinetics showed fast kinetics for the trivalent actinides. As an alternative organic ligand the methylated TODGA derivate Me-TODGA (2-methyl-N,N,Nʹ,Nʹ-tetraoctyldiglycolamide) was tested in combination with the hydrophilic TS-BTPhen. The Am(III) separation was achieved at even higher nitric acid concentrations compared to TODGA.

Abstract

The EURO-GANEX process was developed for co-separating transuranium elements from irradiated nuclear fuels. A hot flow-sheet trial was performed in a counter-current centrifugal contactor setup, using a genuine high active feed solution. Irradiated mixed (carbide, nitride) U80Pu20 fast reactor fuel containing 20 % Pu was thermally treated to oxidise it to the oxide form which was then dissolved in HNO3. From this solution uranium was separated to >99.9 % in a primary solvent extraction cycle using 1.0 mol/L DEHiBA (N,N-di(2-ethylhexyl)isobutyramide in TPH (hydrogenated tetrapropene) as the organic phase. The raffinate solution from this process, containing 10 g/L Pu, was further processed in a second cycle of solvent extraction. In this EURO-GANEX flow-sheet, TRU and fission product lanthanides were firstly co-extracted into a solvent composed of 0.2 mol/L TODGA (N,N,N′,N′-tetra-n-octyl diglycolamide) and 0.5 mol/L DMDOHEMA (N,N′-dimethyl-N,N′-dioctyl-2-(2-hexyloxy-ethyl) malonamide) dissolved in Exxsol D80, separating them from most other fission and corrosion products. Subsequently, the TRU were selectively stripped from the collected loaded solvent using a solution containing 0.055 mol/L SO3-Ph-BTP (2,6-bis(5,6-di(3-sulphophenyl)-1,2,4-triazin-3-yl)pyridine tetrasodium salt) and 1 mol/L AHA (acetohydroxamic acid) in 0.5 mol/L HNO3; lanthanides were finally stripped using 0.01 mol/L HNO3. Approximately 99.9 % of the TRU and less than 0.1 % of the lanthanides were found in the product solution, which also contained the major fractions of Zr and Mo.