A procedure for concentrating americium and curium combined with their separation from plutonium and fission products using a chelating resin

The effect of agitation time and resin quantity on the sorption of americium, curium and europium from mineral acid solutions, using a chelating resin based on aminopolystyrene and Arsenazo I was determined, and the behaviour of plutonium and fission products was investigated under optimum conditions with respect to the sorption of americium and curium. A procedure is proposed for concentrating americium and curium from dilute solutions, combined with their separation from iron, plutonium and fission products. The procedure consists of sorption on the chelating resin from 0.1–1N mineral acid solutions, washing of the resin with 0.5M oxalic acid and 11N sulfuric acid, and elution of americium and curium with 2M triammonium citrate.     

Technetium chemistry in a novel group actinide extraction process

A newly developed method for advanced reprocessing of used nuclear fuel is the Group ActiNide EXtraction (GANEX) process. It is a liquid–liquid extraction process that aims at extracting all the actinides as a group from dissolved used nuclear fuel. This extraction can either be performed after a removal of the bulk uranium or directly on the dissolution liquor. At Chalmers University of Technology in Sweden a solvent that utilizes tributyl-phosphate (TBP) and a molecule from the bis-triazine bipyridine (BTBP) class of ligands dissolved in cyclohexanone has been developed for the use in a GANEX process. Previously the system has not been tested with the presence of technetium that is one of the major fission products. Technetium is often considered a problem within reprocessing since it has a chemical behaviour that differs from most other elements in the spent fuel. Therefore, a special emphasis was put on the investigation of technetium in the selected GANEX system. It was shown that technetium is readily extracted by the GANEX solvent and that cyclohexanone is the main extractant when no other metals were present in the system. It was also found that the presence of uranium decreased the overall technetium extraction despite a slight co-extraction with TBP, while irradiation of the GANEX solvent to large doses (>1 MGy) increased its technetium extraction capability. It was also discovered that an increased nitrate concentration in the aqueous phase and an addition of other fission products both inhibited the technetium extraction even though fission product loading most likely changed the extraction mechanism to co-extraction by BTBP.     

Solubility Thermodynamics of CyMe$$_{}$$-BTBP in Various Diluents Mixed with TBP

The two organic ligands 6,6′-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydrobenzo[1,2,4]triazin-3-yl)[2,2′]bipyridine (CyMe\(_{4}\)-BTBP) and tri-butyl phosphate (TBP) have previously been investigated in different diluents for use within recycling of used nuclear fuel through solvent extraction. The thermodynamic parameters, \(K_{\mathrm{S}}\), \(\Delta C_{p}\), \(\Delta H^{0}\) and \(\Delta S^{0}\), of the CyMe\(_{4}\)-BTBP solubility in three diluents (cyclohexanone, octanol and phenyl trifluoromethyl sulfone) mixed with TBP have been studied at 288, 298 and 308 K, both as pristine solutions and pre-equilibrated with 4 mol\(\cdot \)L\(^{-1}\) nitric acid. In addition, the amount of acid in the organic phase and density change after pre-equilibration have been measured. The solubility of CyMe\(_{4}\)-BTBP increases with an increased temperature in all systems, especially after acid pre-equilibration. This increased CyMe\(_{4}\)-BTBP solubility after pre-equilibration could be explained by acid dissolution into the solvent. Comparing the \(\Delta H^{0}\) and \(\Delta S^{0}\) calculated using \(\Delta C_{p}\) with the same parameters derived from a linear fit indicates temperature independence of all three thermodynamic systems. The change in enthalpy is positive in all solutions.     

Nitric Acid Extraction into TODGA

Advanced solvent extraction processes, namely DIAMEX, SANEX or GANEX, for the separation of the minor actinides (americium and curium) are under development within Europe. The tridentate diglycolamide ligand, TODGA, shows many interesting properties and is under investigation in conjunction with a variety of other extractants for the DIAMEX and SANEX processes as well as the GANEX process. In order to successfully demonstrate these processes, understanding the acid extraction into the organic phases is critical to process flowsheet design and modelling. Here nitric acid extractions into TODGA have been measured and models produced using an equilibrium based approach accounting for nitric acid activities in the aqueous phase.     

Extraction studies of trivalent ions from TALSPEAK medium using phsophonic acid-TBP solvent mixture

Indigenously synthesized extractant, phenyl (octyl) phosphonic acid (POPA) in tri-n-butylphosphate (TBP) and dodecane, has been investigated for the separation of americium from trivalent lanthanides in nitric acid medium as well as diethylene triaminepentaacetic acid (DTPA) and lactic acid mixture (TALSPEAK medium). Various experimental parameters like concentration of DTPA, lactic acid, TBP, nitrate ions and pH of the aqueous feed solution have been optimized to obtain the highest separation factor between americium and europium. Bulk actinide–lanthanide separation reagent, tetra (ethylhexyl) diglycolamide (TEHDGA), was equilibrated with simulated solution of americium and lanthanides, equivalent in concentration to the reprocessing waste originating from PHWR spent fuel. DTPA/lactic acid mixture was used to strip the metal ions from the loaded organic phase and re-extracted into POPA in TBP/dodecane to evaluate the separation factor of individual lanthanides with respect to americium. Very good separation factors between americium and trivalent lanthanides were obtained.     

Properties of PAN-TBP extraction chromatographic material

Three production routes of the preparation of a solid extractant based on tributylphosphate (TBP) embedded in the polyacrylonitrile matrix (PAN) have been studied. The method of direct PAN coagulation with TBP was found to be not viable due to the significant TBP solubility in the coagulation bath. The most suitable PAN-TBP solid extractant was prepared by the well-known impregnation method of ready-made neat PAN beads. The kinetics of uranium extraction from 3 mol L^?1 HNO₃, the effect of nitrate and nitric acids concentrations on the value of weight distribution coefficients D _g as well as the uranium “extraction isotherm” were determined for this material. Uranium extraction was rather fast, approximately 1 h was sufficient for the equilibrium achievement. Capacity for the uranium uptake, measured in batch experiments on PAN-TBP for 0.048 mol L^?1 of uranium in 3 mol L^?1 nitric acid, was found to be q = 0.363 mmol g^?1 (58 % of the theoretical capacity). It was concluded that PAN-TBP material behaves like TBP in liquid–liquid extraction. Extraction capacity determined in column experiments was lower (by about 23 %) than expected from the “extraction isotherm” due to the TBP leaching out of the column. The thus prepared material is therefore not very suitable for multicycle extraction and stripping and can be used once, particularly for the analytical purposes.     

Separation and recovery of Cm from Cm–Pu mixed oxide samples containing Am impurity

Curium was separated and recovered as an oxalate from a Cm–Pu mixed oxide which had been a ²⁴⁴Cm oxide sample prepared more than 40 years ago and the ratio of ²⁴⁴Cm to ²⁴⁰Pu was estimated to 0.2:0.8. Radiochemical analyses of the solution prepared by dissolving the Cm–Pu mixed oxide in nitric acid revealed that the oxide contained about 1 at% of ²⁴³Am impurity. To obtain high purity curium solution, plutonium and americium were removed from the solution by an anion exchange method and by chromatographic separation using tertiary pyridine resin embedded in silica beads with nitric acid/methanol mixed solution, respectively. Curium oxalate, a precursor compound of curium oxide, was prepared from the purified curium solution. 11.9 mg of Cm oxalate having some amounts of impurities, which are ²⁴³Am (5.4 at%) and ²⁴⁰Pu (0.3 at%) was obtained without Am removal procedure. Meanwhile, 12.0 mg of Cm oxalate (99.8 at% over actinides) was obtained with the procedure including Am removals. Both of the obtained Cm oxalate sample were supplied for the syntheses and measurements of the thermochemical properties of curium compounds.     

Separation of Am(III) from nitric acid medium by phosphonium ionic liquid–hydroxyacetamide mixture

A strongly hydrophobic phosphonium ionic liquid, trihexyltet radecylphosphonium bis(trifluoromethanesulfonyl)imide ([P₆₆₆₁₄][NTf₂]) was employed as the diluent for the extraction behavior of Am(III) using N,N-dihexyl-2-hydroxyacetamide(DHHy) as extractant. The extractibility of americium(III) in [P₆₆₆₁₄][NTf₂] phase was measured as a function of various parameters such as aqueous phase acidity (0.1–8 M), extractant concentration (0.01–0.15 M), equilibration time (5–120 min) and temperature (298–333 K). The extraction performance observed in DHHy/[P₆₆₆₁₄][NTf₂] was compared with those observed in N,N-dihexyloctamide (DHOA) in [P₆₆₆₁₄][NTf₂] and DHHy in other diluents such as [C₄mim][NTf₂] and n-dodecane. The effect of temperature on D _Am(III) in ionic liquid system and recovery of Am(III) from the loaded phase were ascertained in detail.     

Extraction of americium(VI) by a neutral phosphonate ligand

Recently the use of the more unusual hexavalent oxidation state of americium has been receiving increased attention for the purpose of developing an efficient Am/Cm or Am/lanthanide separation system. We have already demonstrated the feasibility of performing this separation with 30% TBP in dodecane, and are now looking at different extractants to increase Am(VI) distribution ratios. Following on from this the extraction of bismuth oxidized americium from nitric acid solutions by dibutyl butyl phosphonate has been studied. The results of this study indicate that increasing the basicity of the extractant molecule has significantly improved the extraction efficiency.     

Separation of neptunium, plutonium, americium and curium from uranium with di-(2-ethylhexyl)-phosphoric acid (HDEHP) for radiometric and ICP-MS analysis

The possibility of using di-(2-ethylhexyl)-phosphoric acid (HDEHP) in solvent extraction for the separation of neptunium, plutonium, americium and curium from large amounts of uranium was studied. Neptunium, plutonium, americium and curium (as well as uranium) were extracted from HNO₃, whereafter americium and curium were back-extracted with 5M HNO₃. Thereafter was neptunium back-extracted in 1M HNO₃ containing hydroxylamine hydronitrate. Finally, plutonium was back-extracted in 3M HCl containing Ti(III). The method separates²³⁸Pu from²⁴¹Am for α-spectroscopy. For ICP-MS analysis, the interferences from²³⁸U are eliminated: tailing from²³⁸U, for analysis of²³⁷Np, and the interference of²³⁸UH⁺ for analysis of²³⁹Pu. The method has been used for the analysis of actinides in samples from a spent nuclear fuel leaching and radionuclide transport experiment.     

Sequential separation of neptunium,plutonium, americium and curium using coprecipitation with bismuth phosphate

The reaction of neptunium, plutonium and americium with oxidizing or reducing agents in phosphoric acid solution has been studied to design a separation procedure of the actinide elements using coprecipitation with bismuth phosphate. In the presence of uranium, successive separation of neptunium, plutonium, americium and curium was accomplished by combining the coprecipitation and redox reaction of the elements. The coprecipitation behaviour of fission products during the course of sequential separation of the actinide elements on bismuth phosphate was also discussed.     

Counter-current extraction and separation of U(VI) from a mixture of U(VI)–Th(IV)–Y(III) using tris-2-ethyl hexyl phosphate (TEHP)

Evaluation of tris-2-ethyl hexyl phosphate (TEHP) for counter-current extraction and separation of U(VI) from a mixture of U(VI)–Th(IV)–Y(III) from nitric acid medium was carried out under wide experimental conditions. Batch extraction studies were carried out to investigate the effect of nitric acid concentration in feed solution, U(VI)/Th(IV) ratio and extractant concentration and the results were compared with established solvent such as tri-n-butyl phosphate (TBP) for separation of U(VI) from nitric acid medium. McCabe–Thiele diagrams for extraction as well as stripping of U(VI) were constructed under simulated conditions. Based on batch experiments, six stage counter-current extraction studies were conducted under various TEHP concentration and it was observed that 0.1 M TEHP/n-paraffin was most suitable for selective recovery of U(VI) from a mixture of U(VI)–Th(IV). An optimized condition, 0.1 M TEHP/n-paraffin, 2 M HNO₃ in feed and six number of stages was evaluated for selective extraction and stripping of U(VI) from a solution containing mixture of U(VI)–Th(IV)–Y(III) in nitric acid medium. The U(VI) in strip solution was precipitated using 30 % H₂O₂ at pH ~3. Average particle size of the final precipitate was found to be ~33 μm.     

Gamma Radiolysis of the Highly Selective Ligands CyMe4BTBP and CyMe4BTPhen: Qualitative and Quantitative Investigation of Radiolysis Products

The highly selective nitrogen donor ligands CyMe₄BTBP and CyMe₄BTPhen where γ–irradiated under identical experimental conditions in 1–octanol with and without contact to nitric acid solution. Subsequently, solvent extraction experiments were carried out to evaluate the stability of the extractants against γ–radiation monitoring Am(III) and Eu(III) distribution ratios. Generally, decreasing distribution ratios with increasing absorbed dose were detected for both molecules. Furthermore, qualitative mass spectrometric analyses were performed and ligand concentrations were determined by HPLC-DAD after irradiation to investigate the radiolysis mechanism. An exponential decrease with increasing absorbed dose was observed for both ligands with a faster rate for CyMe₄BTPhen. Main radiolysis products indicated the addition of one or more diluent molecules (1–octanol) to the ligand via prior production of α-hydroxyoctyl radicals from diluent radiolysis. The addition of nitric acid during the irradiation lead to a remarkable stabilization of the system, as the extraction of Am(III) and Eu(III) did not change significantly over the whole examined dose range. Quantification of the remaining ligand concentration on the other hand showed decreasing concentrations with increasing absorbed dose. The stabilization of D values is therefore explained by the formation of 1–octanol addition products which are also able to extract the studied metal ions.     

Tri-iso-amyl phosphate (TAP): An alternative extractant to tri-butyl phosphate (TBP) for reactor fuel reprocessing

Tri-iso-amyl phosphate (TAP), an indigenously prepared extractant was utilized for reactor fuel reprocessing and compared with tri-butyl phosphate (TBP) and tri-n-hexyl phosphate (THP). The potential of these extractants was found to be in the order TAP>THP>TBP by calculating the acid uptake value (K _H). The effect of various parameters such as solvent degradation due to acid hydrolysis, radiation effect, decontamination factor and phase separation were investigated and it was found that TAP was always a better extractant in comparison to THP and TBP. In addition to this, the extraction of fission product contaminants such as ¹⁴⁴Ce, ¹³⁷Cs, ¹⁰⁶Ru, ⁹⁵Zr was almost negligible, even at very high nitric acid concentrations in the aqueous phase, indicating the potential application of TAP in actinide partitioning. Sodium carbonate solution or acidified distilled water was a good strippant for U(VI), similarly, uranium(IV) nitrate stripped Pu(IV) from the organic phase.     

Various flowsheets of actinides recovery with diamides of heterocyclic dicarboxylic acids

Solvents on the base of diamides of heterocyclic dicarboxylic acids are promising alternatives for studied Grouped ActiNide EXtraction (GANEX) solvents. Advantage of these ligands is the possibility of simultaneous extraction not only of residual uranium and plutonium, but also minor actinides—neptunium, americium and curium. Two flowsheets on the base of diamides of heterocyclic dicarboxylic acids for separation of actinides form acidic solutions were developed, tested in laboratory scale and compared. Both flowsheets allow separation of more than 99.95% of actinides from raffinates with high content of lanthanides.     

A novel rapid detection approach for the analysis of radionuclides in environmental samples using graphite MALDI mass spectrometry

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.     

Extractive separation of Th(IV), U(VI), Ti(IV), La(III) and Fe(III) from Zarigan ore

In this study, the effects of various extraction parameters such as extractant types (Cyanex302, Cyanex272, TBP), acid type (nitric, sulfuric, hydrochloric) and their concentrations were studied on the thorium separation efficiency from uranium(VI), titanium(IV), lanthanum(III), iron(III) using Taguchi^??s method. Results showed that, all these variables had significant effects on the selective thorium separation. The optimum separations of thorium from uranium, titanium and iron were achieved by Cyanex302. The aqueous solutions of 0.01 and 1 M nitric acid were found as the best aqueous conditions for separating of thorium from titanium (or iron) and uranium, respectively. The combination of 0.01 M nitric acid and Cyanex272 were found that to be the optimum conditions for the selective separation of thorium from lanthanum. The results also showed that TBP could selectively extract all studied elements into organic phase leaving thorium behind in the aqueous phase. Detailed experiments showed that 0.5 M HNO₃ is the optimum acid concentration for separating of thorium from other elements with acidic extractants such as Cyanex272 and Cyanex302. The two-stage process containing TBP-Cyanex302 was proposed for separation thorium and uranium from Zarigan ore leachate.     

Synergistic extraction of uranium(VI) with tri-n-butyl phosphate and i-butyldodecylsulfoxide

Summary The synergistic extraction of uranium(VI) from aqueous nitric acid solution with a mixture of tri-n-butyl phosphate (TBP) and i-butyldodecylsulfoxide (BDSO) in toluene was investigated. The effects of the concentrations of extractant, nitric acid, sodium nitrate and sodium oxalate on the distribution ratios of uranium(VI) have been studied. The values of enthalpy change for the extraction reactions with BDSO, TBP and a mixture of TBP and BDSO in toluene were -23.2±0.8 kJ/mol, -29.2±1.4 kJ/mol and -30.6±0.6 kJ/mol, respectively. It has been found that the maximum synergistic extraction effect occurs when the molar ratio of TBP to BDSO is close to 1. The composition of the complex of the synergistic extraction is UO₂(NO₃)₂ ^. BDSO ^. TBP.     

Studies on extraction of thorium(IV) from sulfate media with mixtures of long-chain primary amines and tributyl phosphate

The partition of Th(IV) from H₂SO₄ solutions with extractant mixtures of long-chain primary amine (Primene JMT, PrJT) and tributyl phosphate (TBP) is described. Exraction was optimized at PrJT/TBP molar ratio of about 3. The dependence of extraction on acidity, salting agent, extractant concentration, diluent type and temperature, was investigated. From the results obtained, the extracted species are suggested and enthalpy data determined.     

In Situ Biphasic Extractive Fermentation for Hexanoic Acid Production from Sucrose by Megasphaera elsdenii NCIMB 702410

Hexanoic acid production by a bacterium using sucrose as an economic carbon source was studied under conditions in which hexanoic acid was continuously extracted by liquid–liquid extraction. Megasphaera elsdenii NCIMB 702410, selected from five M. elsdenii strains, produced 4.69 g l^?1 hexanoic acid in a basal medium containing sucrose. Production increased to 8.19 g l^?1 when the medium was supplemented by 5 g l^?1 sodium butyrate. A biphasic liquid–liquid extraction system with 10 % (v/v) alamine 336 in oleyl alcohol as a solvent was evaluated in a continuous stirred-tank reactor held at pH 6. Over 90 % (w/w) of the hexanoic acid in a 0.5 M aqueous solution was transferred to the extraction solvent within 10 h. Cell growth was not significantly inhibited by direct contact of the fermentation broth with the extraction solvent. The system produced 28.42 g l^?1 of hexanoic acid from 54.85 g l^?1 of sucrose during 144 h of culture, and 26.52 and 1.90 g l^?1 of hexanoic acid was accumulated in the extraction solvent and the aqueous fermentation broth, respectively. The productivity and yield of hexanoic acid were 0.20 g l^?1 h^?1 and 0.50 g g^?1 sucrose, respectively.