From BTBPs to BTPhens: The Effect of Ligand Pre-Organization on the Extraction Properties of Quadridentate Bis-Triazine Ligands

The synthesis, lanthanide complexation and solvent extraction of An(III) and Ln(III) radiotracers from nitric acid solutions by a pre-organized, phenanthroline-derived bis-triazine ligand CyMe4-BTPhen are described. It was found that the ligand separated Am(III) and Cm(III) from the lanthanides with remarkably high efficiency, high selectivity, and faster extraction kinetics compared to its 2,2’-bipyridine counterpart CyMe4-BTBP. The origins of the ligands extraction properties were established by a combination of solvent extraction experiments, X-ray crystallography, kinetics and surface tension measurements and lanthanide NMR spectroscopy.     

The Extraction of Silver and the Effect of Diluent,Ligand Side Group and Solvent Composition

Solvent extraction and the so called BTBP class of ligands can be used for the separation of the actinides from the rest of used nuclear fuel. One troublesome co-extracting element in this separation is silver.Therefore, two different BTBP molecules, having different side groups have been investigated. It was shown that the silver distribution ratio is higher using the CyMe4-BTBP than theC2 -BTBP ligand. In additional experiments, it was shown that no water soluble silver complex is formed in the CyMe4 system and that the complex is one ligand/metal. No effect of varying the diluent/solvent was proven.     

Solid-phase extraction of plutonium in various oxidation states from simulated groundwater using N-benzoylphenylhydroxylamine

Solid-phase extraction of plutonium in different individual and mixed oxidation states from simulated groundwater (pH 8.5) was studied. The extraction of plutonium species was carried out in a dynamic mode using DIAPAK C16 cartridges modified by N-benzoylphenylhydroxylamine (BPHA). It was shown that the extent of recovery depends on the oxidation state of plutonium. The extraction of Pu(IV) was at the level of 98–99% regardless of the volume and flow-rate of the sample solution. Pu(V) was extracted by 90–95% and 75–80% from 10- and 100-mL aliquots of the samples, respectively, whereas the extraction of Pu(VI) did not exceed 45–50%. An equimolar mixture of Pu(IV), Pu(V), and Pu(VI) was extracted by 74%. The distribution coefficients (K _d) and kinetic exchange capacities (S) of plutonium in various oxidation states were measured. It was found that during the sorption process, Pu(V) was reduced to Pu(IV) by 80–90% after an hour-long contact with the solid phase. Pu(VI) is reduced to Pu(V) by 34% and to Pu(IV) by 55%. In the case of mixed-valent solution of plutonium, only Pu(V) and Pu(IV) were found in the effluents.     

Modeling of Pu(VI) distribution coefficients in 30 % TBP based PUREX process

Hexavalent plutonium (Pu(VI)) is an important solute in the PUREX (plutonium uranium extraction) process. In 30 % TBP based PUREX solvent extraction system, distribution coefficient of Pu(VI) is much lower than that of Pu(IV). This lower distribution coefficient of Pu(VI) may cause unexpected Pu loss during primary HA extraction in low acid flowsheets. An empirical model for Pu(VI) distribution coefficients in 30 % TBP and its temperature dependency has been reported in this paper. Comparison with literature data revealed a reasonably good agreement between the reported experimental and model predicted values.     

Studies on purification of plutonium from silver ions

Precipitation and solvent extraction methods have been investigated for the purification of plutonium from silver from the solution generated during oxidative dissolution of plutonium oxide using Ag(II) ions. Initial experiments have been carried out using thorium as representative of plutonium. Selecting the optimum conditions, the experiments were repeated with plutonium. The results revealed that Pu can be purified from silver ions either by precipitating silver as silver chloride or silver metal followed by Pu(IV) oxalate precipitation or by selective extraction of Pu(IV) into 20% Aliquat-336 or 30% TBP.     

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied. The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.     

Complexes formed between the quadridentate, heterocyclic molecules 6,6'-bis-(5,6-dialkyl-1,2,4-triazin-3-yl)-2,2'-bipyridine (BTBP) and lanthanides(III): implications for the partitioning of actinides(III) and lanthanides(III)

New hydrophobic, tetradentate nitrogen heterocyclic reagents, 6,6'-bis-(5,6-dialkyl-1,2,4-triazin-3-yl)-2,2'-bipyridines (BTBPs) have been synthesised. These reagents form complexes with lanthanides and crystal structures with 11 different lanthanides have been determined. The majority of the structures show the lanthanide to be 10-coordinate with stoichiometry [Ln(BTBP)(NO3)3] although Yb and Lu are 9-coordinate in complexes with stoichiometry [Ln(BTBP)(NO3)2(H2O)](NO3). In these complexes the BTBP ligands are tetradentate and planar with donor nitrogens mutually cisi.e. in the cis, cis, cis conformation. Crystal structures of two free molecules, namely C2-BTBP and CyMe4-BTBP have also been determined and show different conformations described as cis, trans, cis and trans, trans, trans respectively. A NMR titration between lanthanum nitrate and C5-BTBP showed that two different complexes are to be found in solution, namely [La(C5-BTBP)2]3+ and [La(C5-BTBP)(NO3)3]. The BTBPs dissolved in octanol were able to extract Am(III) and Eu(III) from 1 M nitric acid with large separation factors.     

Purification and recovery of plutonium from uranium oxide obtained in the fast reactor fuel reprocessing plant using hydroxyurea as reductant

Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin (HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product obtained can be handled on table top.     

Di(2-ethyl hexyl) butyramide and di(2-ethyl hexyl)isobutyramide as extractants for uranium(VI) and plutonium(IV)

The extraction of uranium(VI) and plutonium(IV) from nitric acid into n-dodecane was studied using two isomeric branched alkyl amides, di(2-ethyl hexyl) butyramide (DEHBA) and di(2-ethyl hexyl) isobutyramide (DEHIBA). The extraction ratios of Pu(IV) at relatively high acidities were higher than the corresponding values for U(VI) in the case of DEHBA. However, with DEHIBA the values for Pu(IV) were negligibly small. Pu(IV) was found to be extracted as trisolvate by DEHBA and as disolvate by DEHIBA. U(VI) was extracted by both the amides. From the study of the extraction reactions at different temperatures, it was shown that all the reactions in the present investigation were enthalpy favoured and entropy disfavoured. Separation of Pu(IV) from bulk of U(VI) was feasible. However, the purity of the separated plutonium was not satisfactory in batch extraction studies.     

Complexation of plutonium(IV) with sulfate at variable temperatures

The complexation of plutonium(IV) with sulfate at variable temperatures has been investigated by solvent extraction method. A NaBrO₃ solution was used as holding oxidant to maintain the plutonium(IV) oxidation state throughout the experiments. The distribution ratio of Pu(IV) between the organic and aqueous phases was found to decrease as the concentrations of sulfate were increased. Stability constants of the 1:1 and 1:2 Pu(IV)-HSO₄ ⁻ complexes, dominant in the aqueous phase, were calculated from the effect of [HSO₄ ⁻] on the distribution ratio. The enthalpy and entropy of complexation were calculated from the stability constants at different temperatures using the Van’t Hoff equation.     

Determination of uranium and plutonium in plutonium based fuels by sequential amperometric titration

A method is described for the sequential determination of uranium and plutonium in plutonium bearing fuel materials. Uranium and plutonium are reduced to U(IV) and Pu(III) with titanous chloride and then titrated with dichromate to two end points which are detected amperometrically using two polarized platinum electrodes. Uranium-plutonium solutions of known concentrations containing plutonium in the proportions of 4, 30, 50, and 70% were analyzed with precisions better than 0.3%, maintaining the amounts of plutonium per aliquot in the range of 2–10 mg. No significant bias could be detected. Several samples of (U, Pu)O₂ and (U, Pu)C were analyzed by this procedure. The effects of iron, fluoride, oxalic acid and mellitic acid on the method were also studied.     

controlled-potential coulometric determination of plutonium with a hydrochloric acid—sulfamic acid electrolyte and phosphate complexing

Total plutonium is selectively determined by a controlled-potential coulometric method in which plutonium is reduced to Pu(III) at 0.25 V (vs. SCE) in a 5.5 M hydrochloric acid—0.015 M sulfamic acid electrolyte, diverse ions are oxidized at 0.57 V, phosphate is added to reduce the Pu(III)—Pu(IV) potential, and Pu(III) is oxidized to Pu(IV) at 0.68 V. None of more than 50 metal ions present in nuclear fuel-cycle material interferes. Many anions are without effect and most interfering ones are removed by preliminary fuming with perchloric acid. The apparatus described consists mainly of commercial components. The relative standard deviation is <0.1% at the 5-mg plutonium level.     

Selective liquid scintillation method of plutonium α-spectrometry

A selective liquid scintillation method is suggested for plutonium determination by alpha-spectrometry in the presence of uranium. The analytical process consists of extracting plutonium from nitric acid solutions using BPHA, TOA and MTOA nitrate as extractants into toluene scintillator. Effect of various extractants and nitric acid concentration on the extraction of Pu(IV) and U(VI) together with scintillations quenching were investigated. A simple and fast two-stage selective scintillation method is also suggested for Pu determination in the irradiated uranioum samples.     

Synthesis of dihydroxyurea and its application to the U/Pu split in the PUREX process

Dihydroxyurea (DHU) was synthesized using tri-associated solid phosgene [bis(trichloromethyl) carbonate] dissolved in dioxane and hydroxylamine hydrochloride dissolved in potassium acetate solution. The reduction of Pu(IV) by DHU was investigated using UV-Vis spectrophotometry. The reduction back-extraction behavior of Pu(IV) in 30% tri-butyl phosphate/kerosene was firstly investigated under conditions of various temperature, various DHU and HNO₃ concentrations and various phase contact times. The results showed that Pu(IV) in the organic phase can be stripped rapidly to the aqueous phase by DHU. Simulating the 1B contactor of the PUREX process using a 0.1 M DHU in 0.36M nitric acid solution as the stripping agent, the separation factors of uranium/plutonium can reach 2.1·10⁴. This indicates that DHU is a promising salt free agent for uranium/plutonium separation.     

Distribution and identification of Plutonium(IV) species in tri-n-butyl phosphate/HNO3 extraction system containing acetohydroxamic acid

There was a significant research progress achieved with the aim to modify conventional PUREX process by stripping of plutonium from the tri-n-butyl phosphate (TBP) extraction product in the form of non-extractable complexes upon addition of back-hold complexation agents. The present paper reports effects of such salt-free complexant, acetohydroxamic acid (HAHA), on distribution ratio of Pu(IV) under wide concentration of nitric acid and additional nitrate. General formula of plutonium species present in the organic phase can be described as Pu(OH)_x(AHA)_y(NO3)_4−x−y·2TBP·wHNO₃.     

Simultaneous determination of uranium and plutonium isotopes in soils by means of single alpha-spectrometry

A simple, rapid and reliable method was developed for the simultaneous determination of uranium and plutonium isotopes by alpha-spectrometry using a single source. A new uranium tracer²³⁰U was applied as well as the²³⁶Pu tracer to determine overall yields of uranium and plutonium isotopes throughout the entire procedure employed. The analytical procedure consists of sample leaching with 8N HNO₃ solution, purification by solvent extraction, simultaneous electrodeposition of U and Pu, and subsequent alpha-spectrometry with a silicon detector. In the solvent extraction using TOA/xylene from 8N HNO₃ solution, the preferential extractability of Pu rather than U permits to purify simultaneously the trace amounts of Pu and the macro amounts of U, as in the case of ordinary soil samples, resulting in favourable peak heights for both isotopes. From a single alpha-spectrum, the determinations of²³⁸U,²³⁴U (and their ratio of²³⁴U/²³⁸U),^239+240Pu, and²³⁸Pu contents were conveniently carried out after correcting the overall yields obtained from²³⁰U and²³⁶Pu activities in the same spectrum. This analytical method was satisfactorily applied to the determination of U and Pu isotope contents in some soils.     

Di(2-ethylhexyl)sulfoxide as an extractant for plutonium(IV) from nitric acid medium

The liquid-liquid extraction behavior of plutonium(IV) from aqueous nitric acid media into n-dodecane by di(2-ethylhexyl)sulfoxide (DEHSO) was investigated over a wide range of conditions. Optimum-parameters such as the aqueous phase acidity, reagent and metal concentrations, etc., were established for efficient extraction-separation of tracer as well as macro levels of plutonium. It was found that the extraction increased with increasing nitric acid concentration up to 6M HNO₃ and then decreased. Extraction also increased with increasing extractant concentration. After loading of the organic phase with 2 to 50 mg/ml of U(VI), extractability of Pu(IV) became considerably lower. Recovery of Pu(IV) from the organic phase was accomplished using dilute uranium(IV) nitrate as the strippant.     

Extraction of plutonium from sulphuric acid by TOA in toluene

The extraction of plutonium(VI) and plutonium(III) from sulphuric acid by TOA in toluene has been studied as a function of the acid and tri-octyl amine concentration. A comparison of the extraction properties of plutonium with those of uranium(VI) and uranium(IV) has been made. It was found that the extraction properties of plutonium(VI) are very similar to those of uranium(VI) and that TOA is a relatively poor extractant for plutonium(III). Uranium(IV) shows better extraction properties than plutonium(III). The results obtained are considered in the light of the stabilities of the complexes formed by these elements in the organic and aqueous phase. A method of separation of both elements by solvent extraction based on changing their oxidation states is suggested.     

Effect of mixed solvent media on the sorption and separation of uranium and plutonium on macroporous resins

Ion-exchange studies on uranium and plutonium using macroporous (MP) anion-exchange resins from an aqueous-organic solvent mixed media were carried out to develop a separation method. Out of the several water miscible organic solvents tried methanol and acetone were found to be best suited. Distribution data were obtained for U(VI) and Pu(IV) using three macroporous resins under various parameters. Based on these data, separation factors for Pu/U were calculated. Column experiments using Tulsion A-27(MP) were also carried out. The method has the advantage of loading plutonium from as low as 1M nitric acid in the presence of methanol or acetone and could be used satisfactorily for its recovery from solutions containing plutonium and uranium.     

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.