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Batch distribution studies show that above ~1 M HNO3Np(IV) and Np(VI) are well extracted into a solvent of 0.2 M TODGA/0.5 M DMDOHEMA in kerosene that has been formulated for the extraction of transuranic actinides in the GANEX (grouped actinide extraction) process. Np(IV) and Np(VI) are quite stable in the solvent phase, although Np(VI) is slowly reduced to Np(IV) on standing. Stripping of Np(IV,VI) ions from the GANEX solvent has been shown to be quite efficient using acetohydroxamic acid providing aqueous HNO3concentrations are below ~0.3 M. In contrast, Np(V) shows much lower distribution ratios and is unstable in the GANEX solvent phase with quite rapid formation of Np(IV) observed. Closer analysis shows this to be due to Np(V) disproportionation, which is enhanced at higher organic phase acidities. Disproportionation of Np(V) is also shown to occur in separate TODGA and DMDOHEMA kerosene solutions. These observations thus enable conditions for neptunium extraction to be optimized during the design of a GANEX flowsheet. [ABSTRACT FROM AUTHOR]

To meet the needs of future closed fuel cycles, the complete recovery of minor actinides, including neptunium, may be required. Neptunium can be fully recovered by modifications to the Plutonium URanium Extraction (PUREX) process but this requires careful control of the Np(V)-(VI) redox reaction in the first solvent extraction contactor to avoid losses to the highly active aqueous raffinate, as occurs in current reprocessing plants. As part of the on-going development of an Advanced PUREX process we report a series of solvent extraction experiments aimed at optimizing neptunium recovery in a process that is based on centrifugal contactors as the extraction equipment. Suitable experimental conditions for Np(V) oxidation were identified through simple stirred 2-phase experiments and single stage mini-centrifugal contactor experiments. A U/Np-active proof-of-principle flowsheet test in a multi-stage centrifugal contactor cascade then demonstrated > 99% extraction of neptunium, thus suggesting the aims for neptunium recovery in advanced fuel cycles can be met by an Advanced PUREX process. [ABSTRACT FROM PUBLISHER]

The accurate prediction of the extraction behaviors of various solutes in PUREX reprocessing process is crucial for the operation and implementation of the actual process. In this paper, an integrated model is developed to predict the extraction behaviors of U, Np, Pu, and HNO3 in the co-decontamination step (1A extraction step) in a pulsed extraction column. The model couples several physical and chemical processes, such as countercurrent flow, mass transfer, and chemical reaction. The mass transfer coefficients and the distribution ratios of U(VI), Pu(IV), Np(IV), Np(V), Np(VI), HNO3 and HNO2 can be obtained using this model. In particular, the redox and disproportion reactions of Np are considered in the model, and the flow direction of Np can be judged under various process conditions and the corresponding influential factors can be analyzed. The judgment is based on the yield calculated from the relative concentration profiles of Np(VI), Np(V), and Np(IV) in the two phases. For neptunium to inter 1AP step, it is necessary to select high nitric acid concentration, low nitrite concentration and especially high flow ratio. For neptunium to inter 1AW step, low nitric acid concentration, high nitrite concentration and low flow ratio are needed. Compared with the experimental data, the relative errors of the distribution ratios of various solutes are less than 30%, and the relative errors of the concentrations of U(VI) and Pu(IV) at the outlet of the organic phase are less than 10%, and our model is indicated to be reliable and applicable for co-decontamination step in the PUREX reprocessing process. [ABSTRACT FROM AUTHOR]

Organic solutions comprising the Actinide Lanthanide Separation Process (ALSEP) solvent consisting of 0.5 M 2-ethylhexyl phosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) and 0.05 M N,N,N',N'-tetra(2-ethylhexyl)diglycolamide (T2EHDGA) in n-dodecane were subjected to low LET and high LET irradiation before and after equilibration with an aqueous phase of 3 M HNO3. Degradation dose constants revealed greater ligand degradation due to gamma irradiation than alpha irradiation for both ligands. Furthermore, equilibration with nitric acid did not have a significant impact on ligand degradation for either irradiation source. Identified degradation products were similar for both gamma and alpha irradiation and occurred mostly through the rupture of the N–Ccarbonyl and C–Oether bonds for T2EHDGA and the C–Oether bond in HEH[EHP]. Acid contact appears to alter the degradation pathway by favoring the formation of higher molecular weight recombination products. Mixed T2EHDGA-HEH[EHP]-NO3 complexes were formed with Nd(III) after extraction from 3 M HNO3, and low LET gamma irradiation of the Nd(III) loaded organic solution produced similar degradation products as the organic solution absent of Nd(III). Interestingly, and likely due to the greater radiolytic susceptibility of T2EHDGA than HEH[EHP], a HEH[EHP]-Nd(III) complex appears to form as the T2EHDGA degrades with increasing absorbed dose. [ABSTRACT FROM AUTHOR]

The diglycolamide 2,2’-oxybis(N,N-di-n-decylpropanamide) (mTDDGA) is being studied as an extractant for actinides and lanthanides in the European Grouped Actinide Extraction (EUROGANEX) process. The aim is the development of a more simplified process using a single extractant instead of a mixture of extractants used in the current EURO-GANEX process. This work presents solvent optimization studies of mTDDGA, with regards to the extraction characteristics of the different diastereomers of mTDGA and of mixed diastereomer solutions. Also radiolysis behavior has been studied by irradiation of solvent extraction systems in a gamma irradiation facility using 60Co. The availability of irradiated organic solutions made it possible to gain valuable insights into the plutonium loading capacity after gammairradiation of the solvent up to 445 kGy and to quantify degradation compounds. Solvent extraction characteristic of the major degradation compounds themselves were determined. Like other methylated diglycolamides, we found a remarkable difference in extraction of up to two orders of magnitude between the two diastereomers. High plutonium loading (36 g L−1) is feasible using this single extractant, even after absorbing a dose of 445 kGy. This remarkable observation is possibly promoted by the presence of the main degradation compound which extracts plutonium verywell. [ABSTRACT FROM AUTHOR]

EURO-GANEX is an innovative hydrometallurgical process for the group separation of transuranic actinides from future spent nuclear fuels. A flowsheet test of a process upset (or ‘maloperation’) has been carried out involving the reduction in the concentration of the scrub acid feed to the extract-scrub contactors. The experimental test was completed in two stages. First, the extractscrub contactors were run under normal flowsheet conditions before initiating the process upset by reducing the scrub acidity from 0.5 to 0.05 mol/L HNO3. At low acidity, a change in the online UV-Vis spectra of the solvent phase indicated the formation of hydrolyzed plutonium species or plutonium colloid formation. Surprisingly, however, this did not lead to the recycle and accumulation of plutonium in the extract-scrub contactors as expected indicating that the EURO-GANEX process is quite robust towards this process upset. Further spectroscopic investigations of the organic-phase species were also performed to characterize the conditions under which the hydrolyzed plutonium species form in the EURO-GANEX solvent as well as separate TODGA and DMDOHEMA phases. The spectroscopic studies supported the view that Pu(IV) hydrolyses at low acidity but the hydrolysis is limited by the organic ligands and is consequently reversible on raising the acidity again. The hydrolysis was also observed in separate TODGA and DMDOHEMA phases. [ABSTRACT FROM AUTHOR]

Liquid–liquid extraction and liquid membrane transport behavior of tetravalent actinide ions viz. Th(IV), Np(IV), and Pu(IV) were investigated for the first time using a diglycolamide (DGA) based dendrimer with a tris(2-aminoethyl)amine (TREN) scaffold as the organic extractant. The generation of 1 dendrimer with six DGA pendent moieties (termed as TREN-G1-DenDGA) extracted Np(IV) more effectively than the other two ions, the trend being Np(IV) > Pu(IV) > Th(IV). The extraction studies of Np(IV) from 3 M HNO3 indicated a 1:1 (metal:ligand) species and the extraction efficiency increased with increasing nitric acid concentration (1–6 M). The transport efficiency of Np(IV) increased with the nitric acid concentration (1–6 M) as well as with the ligand concentration. A very low concentration of 5.75 × 10−4 M ligand, when used as the carrier, resulted in the transport of ca. 25% metal ion transport in 5 h, which increased to >85% with 4.4 × 10−3 M ligand. The transport efficiency of the metal ion across the SLM followed the trend Np(IV) > Th(IV) > Pu(IV). The membrane stability was not satisfactory as seen over a period of 5 days suggesting long-term use may require regular replenishment of the carrier solvent. The effective diffusion coefficient (Deff) of Np(IV)-TREN-G1-DenDGA were determined by the lag-time method and was found to be 5.1 × 10−8 cm2/s. [ABSTRACT FROM AUTHOR]

The production of 238Pu from 237Np, to the specifications required for space flight, requires the separation of plutonium from neptunium and a range of fission products. The valuable neptunium is then recycled into targets for further irradiation. At Oak Ridge National Laboratory (ORNL), a solvent extraction process has been developed to separate neptunium and plutonium that uses sodium nitrite to control the neptunium oxidation. This results in a neptunium product contaminated with sodium, which needs to be removed prior to recycling. We have developed a separation process that results in a sodium-free neptunium product with very low levels of plutonium contamination, there by reducing the number of purification steps and thus simplifying the process. [ABSTRACT FROM AUTHOR]

The Chalmers grouped actinide extraction (CHALMEX) process is focused on the co-separation of actinides from all other elements in spent nuclear fuel solution, with the ultimate purpose of transmuting the actinides into shorter-lived and less radiotoxic elements. Based on solvent extraction equilibrium distribution data of actinides and fission products, a preliminary flowsheet was developed and tested in batch mode. The flowsheet consists of one extraction step with the CHALMEX FS-13 solvent (25 mM CyMe4-BTBP in 30% v/v TBP and 70% v/v FS-13), using hydrophilic masking agents (20 mM bimet and 0.2 M mannitol) in the aqueous phase for the complexation of troublesome fission products. Two nitric acid scrub steps (0.5 M HNO3) were efficient in removing co-extracted acid, all molybdenum and the majority of silver. Two stripping stages (0.5 M glycolic acid at pH 4) were efficient in recovery of the actinides from the organic phase. The need for a solvent clean-up stage for the removal of nickel, cadmium, iron and the remaining silver from the organic phase was demonstrated. Based on the distribution data, it was calculated that a 99.9% recovery of americium is possible using only 3 ideal extraction stages, 3 ideal scrubbing stages and 2 ideal stripping stages. [ABSTRACT FROM AUTHOR]

The structure of complex organic solutions based on N,N,N',N'-tetraoctyldiglycolamide (TODGA) used in actinide separation processes were characterized after water and neodymium extraction. The organic phases were characterized using FTIR spectroscopy, ESI-MS spectrometry, Small and Wide Angle X-ray Scattering (SWAXS) and molecular dynamic simulations. The role of the "phase modifier" like 1-octanol (used in the innovative SANEX process) or the "co-extractant" like DMDOHEMA (used, for example, in the EURO-GANEX process) added to the organic phase to avoid third-phase formation was investigated at molecular and supramolecular scales. Without metallic cations in 0.2 mol/L TODGA solution in n-heptane, TODGA molecules are organized in monomers and small aggregates containing mainly two or three TODGA molecules. After Nd(NO3)3 extraction, complexes or aggregates are formed in which each neodymium cation is surrounded in average by 2 to 4 TODGA, 3 nitrate ions and 1 to 2 water molecules. Polymetallic species (aggregates) may also be formed. Adding 5%Vol 1-octanol or 0.5 mol/L DMDOHEMA to 0.2 mol/L TODGA solution in n-heptane leads to the formation of mixed aggregates containing TODGA and octanol molecules or TODGA and DMDOHEMA molecules. These molecules both interact directly with the extracted metallic cation in the organic phase, modifying the structure of the neodymium species, and improving their solubility in this organic phase. At least one TODGA molecule remains in the cation first coordination sphere in both cases and preserves its significant role in the complexation/selectivity of the cations. But it is clear that both 1-octanol and DMDOHEMA have a role in the speciation of the cations in the organic phase: 1-octanol replaces water molecules in the first coordination sphere of Nd, acting as a co-surfactant; DMDOHEMA drives the organization and DMDOHEMA molecules partially replace TODGA in the first coordination sphere of Nd. [ABSTRACT FROM AUTHOR]

Studies have been performed with the purpose of determining the optimal solvent composition of a Chalmers grouped actinide extraction (CHALMEX) solvent for the selective co-extraction of transuranic elements in a novel Grouped ActiNide EXtraction (GANEX) process. The solvent is composed of 6,6'-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-[1,2,4]-triazin-3-yl)-[2,2']-bipyridine (CyMe4-BTBP) and tri-n-butyl phosphate (TBP) in phenyl trifluoromethyl sulfone (FS-13). The performance of the system has been shown to significantly depend on the ratios of the two extracting agents and the diluent to one another. Furthermore, the performance of the determined optimal solvent (10 mM CyMe4-BTBP in 30% v/v TBP and 70% v/v FS-13) on various simulated PUREX raffinate solutions was tested. It was found that the solvent extracts all transuranic elements with high efficiency and good selectivity with regard to most other elements (fission products/activation products) present in the simulated PUREX raffinate solutions. Moreover, the solvent was found to extract a significant amount of acid. Palladium, silver, and cadmium were co-extracted along with the TRU-radionuclides, which has also been observed in other similar CHALMEX systems. The extraction of plutonium and uranium was preserved for all tested simulated PUREX raffinate solutions compared to experiments using trace amounts. [ABSTRACT FROM AUTHOR]

Liquid-liquid distribution data were determined for the extraction of nitric acid, Am(III), Cm(III), and lanthanides(III) from 0.1–7 mol/L nitric acid into 0.5–1 mol/L N,N'-dimethyl-N,N'-dioctyl-2-[2-(hexyloxy)ethyl]-malonamide (DMDOHEMA) dissolved in kerosene. Nitric acid extraction was accurately modeled accounting for the adducts (HNO3)L2, (HNO3)L, and (HNO3)2L. To model Am(III) extraction, the following complexes were taken into account, Am(NO3)3L4, Am(NO3)3(HNO3)L3, and a third complex, Am(NO3)3(HNO3)2L3 (considering spectroscopic results) or Am(NO3)3(HNO3)2L2 (considering slope-analysis results). Separation factors for Am(III) over Cm(III) and the lighter lanthanides(III) are practically independent of nitric acid concentration. Am(III)/Ln(III) separation factors decrease with increasing nitric acid concentration for the heavier lanthanides(III). [ABSTRACT FROM AUTHOR]

A new N-atom donor extractant 2,9-bis(1-(2-ethylhexyl)-1H-1,2,3-triazol-4-yl)-1,10-phenanthroline (EH-BTzPhen) was synthesized and used in solvent extraction studies to separate the trivalent minor actinide americium(III) and curium(III) from europium(III), representing fission product lanthanides. The extractant was found to be soluble in 1-octanol and in the Aliquat-336 nitrate ([A336][NO3]) ionic liquid diluent, but insoluble in n-dodecane. The [A336][NO3] is a fully incinerable room temperature ionic liquid, and has a higher flash point than aliphatic diluents such as 1-octanol. The EH-BTzPhen proved to be effective for selective minor actinide extraction only in combination with 2-bromohexanoic acid synergist when 1-octanol was used as a diluent (SFAm/Eu > 200). The change of the diluent from 1-octanol to Aliquat-336 nitrate allowed selective An(III) extraction from low acidity feed solutions without the need of a synergist (SFAm/Eu ~ 70 and SFAm/Cm ~ 1.9–2.2). The phase transfer kinetics of the ligand-metal complexes is however very slow at 22 °C in the case of both solvents. With this newly synthesized extractant, the achieved SFAm/Cm was comparable to the values achieved with the established CyMe4BTBP and CyMe4BTPhen extractants. [ABSTRACT FROM AUTHOR]

Extraction of Np(IV), Pu(IV), U(VI), and Np(VI) was studied using N,N-di-2-ethylhexyl derivatives iso-butyramide (D2EHIBA) and n-propanamide (D2EHPrA). D2EHIBA showed higher selectivity for U(VI) over Np(IV) and Pu(IV). A flowsheet for the separation of Np(IV) from a mixture of uranium, neptunium, and plutonium in 3 M HNO3 was developed exploiting the selectivity of D2EHIBA for U(VI) over the tetravalent actinides followed by the selective stripping of Np(IV) from the di-n-hexyl-octanamide (DHOA) medium using acetohydroxamic acid (AHA). The results indicated that >90% of the neptunium present in the mixture could be recovered at the final step with a decontamination factor >103. [ABSTRACT FROM AUTHOR]

The feasibility of separating americium from curium and the lanthanides was studied in batch solvent extraction experiments using an organic solvent composed of 0.01 mol L−1 2,9-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-1,2,4-benzotriazin-3-yl)-1,10-phenanthroline (CyMe4BTPhen) in the ionic liquid Aliquat-336 nitrate [A336][NO3] from feed solutions containing N,N,N',N'-tetraethyl diglycolamide (TEDGA), stable lanthanide ions and tracers of 241Am, 244Cm, and 152Eu. The combined use of a lipophilic and a hydrophilic ligand with opposite selectivity for Am(III) vs. Cm(III) and Ln(III) allowed the separation of Am(III) from Cm(III) and Ln(III) from moderate acidic feed solutions (1 mol L−1 HNO3, SFAm/Cm = 3.1–3.9, SFAm/La > 75, SFAm/Eu ≥ 3000). [ABSTRACT FROM AUTHOR]

Several combinations of lipophilic extractants have been studied with the objective of investigating the synergistic behavior of metal ion extraction and its relation to reverse micelle formation. Combinations of acidic and neutral extracting reagents that have displayed synergic behavior in other studies were tested. While most of these either showed limited synergy or no correlation with water uptake under the conditions used here, two of the systems showed extraction synergy. These two extraction systems: 2-bromodecanoic acid with 2,2ʹ:6ʹ,2ʺ-terpyridine; and tributyl phosphine oxide with 2-thenoyltrifluoroacetone were modeled based on two different theoretical approaches: microemulsion-assisted extraction; or the formation of mixed complexes in the organic phase. Comparisons between the two models showed that in the case of bromodecanoic acid and terpyridine the approach of mixed complexes provided a somewhat better fit while the second system was not modeled well with either approach. Further work on the species present in the organic phase is needed to determine the role, if any, of aggregates in synergistic solvent extraction systems. [ABSTRACT FROM AUTHOR]

Distribution data for the partition of nitric acid between nitric acid solution and a solvent phase comprising various combinations of TODGA, octanol, and inert kerosene diluent have been generated, covering a range of conditions from 0 to 9 mol/L HNO3(aq), 0–100% octanol, 0–0.4 mol/L TODGA over a temperature range from 10°C to 50°C. The data have been used to derive a model describing the nitric acid equilibrium between the phases suitable for incorporation in process models of the innovative SANEX process, for example. For the nitric acid/octanol/diluent system, it was found that an accurate prediction of nitric acid distribution could be achieved using a model allowing 1:1, 1:2, and 1:3 nitric acid/octanol adducts. For the nitric acid/TODGA/diluent system, the best models were found to be those allowing 4:1, 3:1, 2:1, 1:1, and 2:2 nitric acid/TODGA adducts. Superimposing the models for nitric acid distribution into the individual extractants and comparing with experimental results for the nitric acid/octanol/TODGA system showed systematic differences indicative of antagonistic and synergistic effects applying in the ranges 0.5–1.5 mol/L HNO3 and >1.5 mol/L HNO3, respectively. These effects were modelled by the inclusion of 0:1:2, 1:1:1, 2:1:3, and 3:1:2 nitric acid/TODGA/octanol adducts. The effect of temperature on nitric acid extraction was well described by an Arrhenius type expression with an activation energy of −25.7 kJ/mol. No diluent dependence was found for nitric acid extraction. [ABSTRACT FROM AUTHOR]

The applicability of 2,6-bis[1-(propan-1-ol)-1,2,3-triazol-4-yl)]pyridine (PyTri-Diol) in the innovative Selective ActiNide EXtraction (i-SANEX) process as water-soluble complexing agent was studied. Preliminary batch experiments were aimed at identifying the optimal formulation of the PyTri-Diol solution and at preparing the ground for single-stage centrifugal contactor experiments. A N,N,N′,N′ tetraoctyl diglycolamide (TODGA)-based solvent loaded using a spiked synthetic Plutonium and Uranium Reduction EXtraction raffinate and the optimized PyTri-Diol aqueous phase were contacted in a single-stage annular centrifugal contactor setup with three different flow-rate conditions. No hydrodynamic problems were encountered, and promising minor actinide separation from other cations was achieved with satisfactory kinetics and stage efficiency. The flow sheet of a TODGA-PyTri-Diol-based i-SANEX process was designed exploiting batch and single-stage data, promoting the CHON compliant PyTri-Diol as excellent alternative to the formerly used SO3-Ph-BTP. [ABSTRACT FROM AUTHOR]

A Grouped ActiNide EXtraction (GANEX) process for the extraction of actinides from used nuclear fuel for transmutation purposes has been investigated. The studied solvent consists of phenyl trifluoromethyl sulfone (FS-13), CyMe4-BTBP, and TBP, a combination that has previously shown promising results. The time to reach extraction equilibrium for the system has been found to be less than 20 min. A 2:1 complex has been found between CyMe4-BTBP and americium(III) or curium(III), whereas plutonium(IV) and CyMe4-BTBP create a 1:1 complex. The extraction of fission product is low in the system. [ABSTRACT FROM AUTHOR]

Over the last decade there has been much interest in the applications of diglycolamide (DGA) ligands for the extraction of the trivalent lanthanide and actinide ions from PUREX high active raffinates or dissolved spent nuclear fuel. Of the DGAs, the N,N,N',N'-tetraoctyldiglycolamide (TODGA) is the best known and most widely studied. A number of new actinide separation processes have been proposed based on extraction with TODGA. This review covers TODGAbased processes and extraction data, specifically focusing on how phase modifiers have been used to increase metal loading and thus enhance the operating process envelopes. Effects of third phase formation and the organic phase speciation are reviewed in this context. Relevant aspects of the extraction chemistry of important solvents (TODGA-modifier-diluent combinations) are described and their performances demonstrated by a consideration of the published flowsheet tests. It is seen that modifiers are successfully enabling the use of TODGA in actinide separation processes but to date the identification and testing of suitable modifiers has been rather empirical. There is a growing understanding of the fundamental chemistry occurring in the organic phase and how that affects extractant speciation andmetal loading capacity but studies are still needed if TODGA-based flowsheets are to become an industrially deployable option for minor actinide (MA) recovery processes. [ABSTRACT FROM AUTHOR]

The radiolytic degradation of di-2-ethylhexyl-isobutyramide (DEHiBA) was examined by subjecting the compound to gamma irradiation, measuring the remaining concentration of the intact compound, identifying the degradation products, and measuring uranium distribution ratios. The combined effects of radiation dose, contact with aqueous solutions of HNO3, and aeration were also examined. The DEHiBA displayed significant stability at doses up to 1000 kGy, undergoing a slow exponential concentration decrease that was accompanied by the appearance of multiple degradation products. The most abundant compounds that were formed by radiolysis resulted from cleavage of the Ccarbonyl-N and Cethylhexyl-N bonds, generating di-ethylhexylamine and mono-ethylhexyl-isobutyramide. Acid contact did alter the radiolytic pathways, with acid favoring cleavage of the Ccarbonyl-N bond, while a more diverse array of compounds was formed in the absence of acid. Pulsed radiolysis experiments were also conducted, in which picosecond bursts of energetic electrons were used to irradiate solutions of dodecane containing DEHiBA; formation of the dodecane radical cation was implicated, which serially reacted with DEHiBA to form a radical or radical cation species intermediate in the formation of the observed products. The slow degradation kinetics suggests that DEHiBA possesses good potential for selective extraction of uranium in fuel cycle extraction operations. [ABSTRACT FROM AUTHOR]

An Advanced TALSPEAK (trivalent actinide–lanthanide separations by phosphorus-reagent extraction from aqueous complexes) counter-current flowsheet test was demonstrated using a simulated feed spiked with radionuclides in annular centrifugal contactors. A solvent comprising 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP] or PC88A) inn-dodecane was used to extract trivalent lanthanides away from the trivalent actinides Am3+and Cm3+, which were preferentially complexed in a citrate-buffered aqueous phase withN-(2-hydroxyethyl)ethylenediamine-N,N´,N´-triacetic acid (HEDTA). In a 24-stage demonstration test, the trivalent actinides were efficiently separated from the trivalent lanthanides with decontamination factors >1000, demonstrating the excellent performance of the chemical system. Clean actinide and lanthanide product fractions and spent solvent with very low contaminations were obtained. The results of the process test are presented and discussed. [ABSTRACT FROM PUBLISHER]

A solvent extraction system was developed for separating trivalent actinides from lanthanides. This “Advanced TALSPEAK” system uses 2-ethylhexylphosphonic acid mono-2-ethylhexyl ester (HEH[EHP]) to extract the lanthanides into ann-dodecane-based solvent; the actinides are retained in a citrate-buffered aqueous phase by complexation to a polyaminocarboxylate ligand. Several aqueous-phase ligands were investigated, andN-(2-hydroxyethyl)ethylenediamine-N,N’,N’-triacetic acid (HEDTA) was chosen for further study. Batch distribution measurements indicate that the separation of americium (Am) from the light lanthanides increases as the pH increases. However, previous investigations indicated that the extraction rates for the heavier lanthanides decrease with increasing pH. Therefore, a balance between these competing effects is required. An aqueous phase at pH 2.6 was chosen for further process development, because this offered optimal separation. Centrifugal-contactor single-stage efficiencies were measured to characterize the system’s performance under flow conditions, and an Advanced TALSPEAK flowsheet was designed. [ABSTRACT FROM PUBLISHER]

Routing neptunium to a single product in spent nuclear fuel reprocessing is a significant challenge. In this work, we have further improved the simulation of neptunium extraction in an advanced PUREX flowsheet by applying a revised model of the Np(V)–Np(VI) redox reaction kinetics, a new nitric acid radiolysis model, and by evaluating various models for the nitrous acid distribution coefficient. The Np disproportionation reaction is shown to have a negligible effect. The models are validated against published ‘cold test’ experimental results; the ‘hot test’ simulation suggests that high neptunium radiolysis could help to achieve high recoveries using this flowsheet. [ABSTRACT FROM AUTHOR]

An Advanced PUREX process has been developed for separation and recycling of neptunium from spent nuclear fuel. This work presents a new flowsheet simulation model for the extraction of neptunium using centrifugal contactors, where mass transfer is modeled using two-film theory and a linear driving force. Distribution coefficients and neptunium redox reactions are modeled using published models. Mass transfer between the organic and aqueous phases in the phase separation zone is shown to have a negligible effect. The model is applied to a previously tested flowsheet and its predictions are shown to be in good agreement with the experimental results. [ABSTRACT FROM PUBLISHER]

The tridentate diglycolamide ligandN,N,N′,N′-tetraoctyl diglycolamide (TODGA) shows many interesting properties and is a very good extractant for the minor actinides (MAs) and lanthanides but, due to its low loading capacity, requires a phase modifier when used in a solvent extraction process. Consequently, applications of TODGA in conjunction with tri-butyl phosphate (TBP) in novel DIAMEX and SANEX processes for recovering MAs have been reported. However, TODGA and TBP also extract nitric acid and this has a significant influence on process performance. Here new distribution data for the extraction of nitric acid into solvent phases containing TODGA and TBP have been collected and modeled using an equilibrium-based approach accounting for nitric acid activities in the aqueous phase. Models for the extraction of nitric acid using the individual extractants were obtained using a variety of complexes and these were then combined to give the extraction of the mixed TODGA and TBP solvent. Using this approach, the nitric acid extraction of the mixed TODGA/TBP system can be reliably reproduced indicating that no significant synergistic or antagonistic complexes are formed in solution. [ABSTRACT FROM PUBLISHER]

An americium(III) selective separation procedure was developed based on the coextraction of trivalent actinides (An(III)) and lanthanides (Ln(III)) by TODGA (N,N,N′,N′-tetraoctyl-diglycolamide), followed by Am(III) selective stripping using the hydrophilic complexing agent TS-BTPhen (3,3′,3ʺ,3ʺ′-[3-(1,10-phenanthroline-2,9-diyl)-1,2,4-triazine-5,5,6,6-tetrayl]tetrabenzenesulfonic acid). Distribution ratios were found at an acidity of 0.65 mol L−1nitric acid that allowed for the separation of Am(III) from Cm(III) (DCm> 1; DAm< 1), giving a separation factor between curium and americium of SFCm/Am= 3.6 within the stripping step. Furthermore, Am(III) was readily separated from the lanthanides with the lowest selectivity for the Ln(III)/Am(III) separation being lanthanum with a separation factor of SFLa/Am= 20. The influence of the TS-BTPhen concentration on Am(III) distribution ratios was studied, giving a slope (logD vs. log[TS-BTPhen]) of approximately −1 for the stripping of An(III) with TS-BTPhen from the TODGA-based organic phase. Time-resolved laser fluorescence spectroscopy (TRLFS) measurements of curium(III) were used to analyze the speciation of Cm(III)-TS-BTPhen complexes. Both 1:1 and 1:2 complexes were identified in single-phase experiments. The formation of the 1:1 complex was suppressed in 0.5 mol L−1nitric acid but it was significantly present in HClO4at pH 3. Conditional stability constants of the complex species were calculated from the TRLFS experiments. [ABSTRACT FROM PUBLISHER]

The separation of actinides(III) from used nuclear fuel is a key step for the recycling of used nuclear fuel in innovative fuel cycles. However, high neutron dose rates and heat load of short-lived curium isotopes complicate the production and handling of new nuclear fuel containing curium(III) and americium(III). Therefore, new processes have to be developed separating only americium(III) from PUREX (Plutonium-Uranium Recovery by Extraction) raffinate. This is achieved by coextracting An(III) and Ln(III) from PUREX raffinate usingN,N,N’,N’-tetraoctyl-diglycolamide (TODGA), followed by the subsequent selective stripping of Am(III) by a water-soluble bis-triazinyl-bipyridin (sodium 3,3’,3’’,3’’’-([2,2’-bipyridine]-6,6’-diylbis(1,2,4-triazine-3,5,6-triyl))tetrabenzenesulfonate, SO3-Ph-BTBP). The selectivity for Am(III) over Cm(III) isSFCm(III)/Am(III)≈ 2.5, due to the inverse selectivity of TODGA and SO3-Ph-BTBP. Additionally, a separation factor up to 1200 is achieved for the separation of Eu(III) from Am(III). The results demonstrate that the presented system works very well even at acidic conditions using nitric acid as the aqueous phase and does not require additional salting out agents. [ABSTRACT FROM PUBLISHER]

The stability of solvents containing the C5-BPP (2,6-Bis(5-(2,2-dimethylpropyl)-1H-pyrazol-3-yl)pyridine) extractant against gamma-irradiation up to 330 kGy adsorbed dose was evaluated using HPLC mass spectrometric and solvent extraction methods. The main radiolysis products were identified using high-resolution mass spectrometry. Quantitative analysis of the residual C5-BPP concentration in irradiated samples was accomplished by a specifically developed MRM (Multiple Reaction Monitoring) method. A mixture of C5-BPP with 2-bromohexanoic acid in kerosene and 1-octanol, which had been used in an actinide(III)/lanthanide(III) separation process demonstration test, was also studied for its radiolytic stability and a remarkable protective effect of the presence of nitric acid during irradiation was found. [ABSTRACT FROM AUTHOR]

Iron oxide (Fe3O4) particles functionalized with diglycolamic acid (Fe-DGAH) were synthesized and characterized by TG-DTA, X-Ray diffraction,1H NMR, and scanning electron microscopy (SEM). The extraction behavior of Am(III) and Eu(III) in Fe-DGAH was studied from dilute nitric acid medium to examine the feasibility for the mutual separation of trivalent actinides and lanthanides using Fe-DGAH. For this purpose, the effect of various parameters such as the duration of equilibration and concentrations of europium, nitric acid, and diethylenetriaminepentaacetic acid (DTPA) in the aqueous phase on the distribution ratio (Kd) of Am(III) and Eu(III) was studied. The conditions needed for the efficient separation of Am(III) from Eu(III) were optimized using DTPA. The distribution ratio of ˜104mL/g was obtained for both Am(III) and Eu(III) at pH 3, and it decreased with an increase in the concentration of nitric acid. However, a separation factor of Eu(III) over Am(III) of ˜150 was achieved in the presence of DTPA. Rapid sorption of metal ions in the initial stages of equilibration followed by the establishment of equilibrium occurred within 2 h. The sorption data were fitted to the Langmuir adsorption model, and the apparent europium sorption capacity was determined to be ˜50 mg/g. The study indicated the feasibility of using Fe-DGAH particles for magnetic separation of Eu(III) from Am(III) with high separation factors. [ABSTRACT FROM AUTHOR]

High-level liquid waste (HLLW) produced from the reprocessing of the spent nuclear fuel still contains moderate amounts of uranium, transuranium (TRU) actinides, and fission products, and thus constitutes a permanent hazard to the environment. The partitioning and transmutation (P&T) strategy has increasingly attracted interest for the safe treatment and disposal of HLLW, in which the partitioning of HLLW is one of the critical technical issues. Two improved trialkylphosphine oxide (TRPO) processes for the removal of actinides have been developed to treat Chinese HLLW, based on the original TRPO process. In one improved processN,N-diethylhydroxylamine as a reducing agent was used for reducing Np(V) and Np(VI) to Np(IV) in order to improve the extraction efficiency of Np. In the other improved process, ammonium vanadate as an oxidizing agent was used for oxidizing Np(V) and Np(IV) to Np(VI) in order to improve the extraction efficiency of Np. Radioactive tracer tests of two improved TRPO processes were carried out using 30-stage 10-mm-diam annular centrifugal contactors and simulated HLLW containing U, Np, Pu, and Am. The test results showed that the decontamination factor of total α activity was >1 × 105. During the test, 30-stage 10-mm-diam annular centrifugal contactors worked in a stable manner continuously, with no stage failing or any interruption of the operation. [ABSTRACT FROM PUBLISHER]

In this paper the development and laboratory-scale demonstration of a novel “innovative-SANEX” (SelectiveActinideExtraction) process using annular centrifugal contactors is presented. In this strategy, a solvent comprising theN,N,N’,N’-tetraoctyldiglycolamide (TODGA) extractant with addition of 5 vol.-% 1-octanol showed very good extraction efficiency of Am(III) and Cm(III) together with the trivalent lanthanides (Ln(III)) from simulatedPlutoniumUraniumRefining byExtraction (PUREX) raffinate solution without 3rdphase formation. Cyclohexanediaminetetraacetic acid (CDTA) was used as masking agent to prevent the co-extraction of Zr and Pd. An(III) and Ln(III) were co-extracted from simulated PUREX raffinate, and the loaded solvent was subjected to several stripping steps. The An(III) were selectively stripped using the hydrophilic complexing agent SO3-Ph-BTP (2,6-bis(5,6-di(sulfophenyl)-1,2,4-triazin-3-yl)pyridine). For the subsequent stripping of the Ln(III), a citric acid based solution was used. A 32-stage process flow-sheet was designed using computer-code calculations and tested in annular miniature centrifugal contactors in counter-current mode. The innovative SANEX process showed excellent performance for the recovery of An(III) from simulated High Active Raffinate (HAR) solution and separation from the fission and activation products. ≥ 99.8% An(III) were recovered with only low impurities (0.4% Ru, 0.3% Sr, 0.1% Ln(III)). The separation from the Ln(III) was excellent and the Ln(III) were efficiently stripped by the citrate-based stripping solution. The only major contaminant in the spent solvent was Ru, with 14.7% of the initial amount being found in the spent solvent. Solvent cleaning and recycling therefore has to be further investigated. This successful spiked test demonstrated the possibility of separating An(III) directly from HAR solution in a single cycle which is a great improvement over the former multi-cycle strategy. The results of this test are presented and discussed. [ABSTRACT FROM AUTHOR]

A flowsheet for a novel GANEX (GroupedActiNideEXtraction) process has been tested in a spiked flowsheet trial in a 32 stage plutonium-active centrifugal contactor rig with a simulant feed that contained 10 g/L plutonium as well as some fission products and other transuranic actinides. The solvent system used was a combination of 0.2 mol/L N,N,N’,N’-tetraoctyl diglycolamide (TODGA) and 0.5 mol/L N,N’-(dimethyl-N,N’-dioctylhexylethoxy-malonamide (DMDOHEMA) in a kerosene diluent that co-extracted actinides and lanthanides. Actinides were subsequently selectively co-stripped away from the lanthanides using a sulphonated and, therefore, hydrophilic bis-triazinyl pyridine (BTP) complexant in conjunction with acetohydroxamic acid (AHA). Plutonium and americium recoveries were high with decontamination factors across the strip contactors of ˜14,000 and ˜390, respectively. However, approximately 30% of neptunium was lost to the aqueous raffinate which was due to recycling within the first extract-scrub section causing a large build-up of neptunium. Some accumulation of strontium was also observed but in this case it was fully directed to the raffinate stream. In the stripping section, a small fraction of europium (taken as a model lanthanide ion),ca. 7%, was found in the actinide product stream. Modelling of selected data using the PAREX code has shown that even with a relatively simplistic treatment, reasonable agreement between modelling and experiment can be obtained, giving confidence in the use of modelling to refine the GANEX flowsheet design prior to further testing with irradiated fast reactor fuel. [ABSTRACT FROM AUTHOR]

The direct selective separation of the trivalent actinides americium and curium from a simulated Plutonium Uranium Refining by EXtraction (PUREX) raffinate solution by a continuous counter-current solvent extraction process using miniature annular centrifugal contactors was demonstrated on a laboratory scale. In a 32-stage spiked test (12 stages for extraction, 16 stages for scrubbing, and 4 stages for Am/Cm stripping), an extractant mixture of CyMe4BTBP and TODGA in a TPH/1-octanol mixture was used. The co-extraction of some fission and corrosion product elements, such as zirconium and molybdenum, was prevented by using oxalic acid. Co-extracted palladium was selectively stripped using an L-cysteine scrubbing solution and the trivalent actinides were selectively stripped using a glycolic acid-based stripping solution. It was demonstrated that a selective extraction and high recovery of > 99.4% of the trivalent minor actinides was achieved with low contamination by fission and corrosion products. The product contained 99.8% of the initial americium and 99.4% of the initial curium content. The spent solvent still contained high concentrations of Cu, Cd, and Ni. The experimental steady-state concentration profiles of important solutes were determined and compared with those from computer-code calculations. [ABSTRACT FROM AUTHOR]