Extraction Of Plutonium Research Articles

Thermodynamic modeling of Pu(IV) and nitric acid extraction by 1.1 M tri-iso-amyl phosphate in n-dodecane

A thermodynamic model for the liquid–liquid extraction of tetravalent plutonium (Pu(IV)) by 1.1 M Tri iso Amyl Phosphate (TiAP) in n-dodecane was developed. The activity coefficients of species in the aqueous phase were estimated using eUNIQUAC activity coefficient model. The plutonium speciation in the aqueous nitric acid solution of 1–6 M concentration region was modelled by taking account of mono nitrate (Pu(NO3)3+), di nitrate $$({\text{Pu}}({\text{NO}}_{3} )_{2}^{2 + } )$$ , tetra nitrate (Pu(NO3)4) and hexa nitrate $$({\text{Pu}}({\text{NO}}_{3} )_{6}^{2 - } )$$ complexes. The eUNIQUAC activity coefficient model parameter and stability constant of tetra and hexa nitrate complexes were determined using the experimental water activity as well as extraction data. The distribution coefficient of nitric acid and Pu(IV) in the 1.1 M TiAP was estimated by accounting HNO3·TiAP, HNO3·2TiAP and Pu(NO3)4·2TiAP complexes in the organic phase. A good correlation between the estimated and experimental organic phase concentration of nitric acid and plutonium nitrate was obtained.

Batch Tests for Optimisation of Solvent Composition and Process Flexibility of the CHALMEX FS-13 Process

ABSTRACT 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.

Neptunium extraction by N,N-dialkylamides

Abstract Separation of neptunium by solvent extraction has been based on tributylphosphate (TBP) for decades, but TBP is not fully incinerable, which adds to the burden of long-lived radioactive waste. Alternatives to TBP for uranium and plutonium extraction, such as the N,N-diakylamides, previously have been explored in the hopes of transitioning to an extractant that is incinerable. Four N,N-diakylamides, N,N-dihexylhexanamide (DHHA), N,N-dihexyloctanamide (DHOA), N,N-di(2-ethylhexyl)butanamide (DEHBA), and N,N-di(2-ethylhexyl)-iso-butanamide (DEHiBA) were considered in this work for their potential to extract millimolar concentrations of Np(IV), Np(V), and Np(VI) from nitric acid solutions into organic solutions containing 1 M extractant in Exxsol D60. Under these conditions the branching of the alkyl substituents affects the extractability of Np(VI) and Np(IV), causing three of the dialkylamides, DHHA, DHOA and DEHBA, to extract neptunium in the expected order Np(VI) > Np(IV) > > Np(V). In contrast, branched DEHiBA is so poor an extractant for Np(IV) that the extraction order becomes Np(VI) > > Np(V) > Np(IV) between 0.1 and 5.6 M HNO3 due to partial oxidation of the Np(V) in nitric acid.

Radiation Hardness of Vinyl-Pyridinium Anion Exchanger of AXIONIT VPA-2 Brand in Nitrate Solutions

Radiation hardness of vinyl-pyridinium anion exchanger of AXIONIT VPA-2 brand, synthesized at Axion–Rare and Precious Metals AO (Perm) in the process of sorption extraction of plutonium from nitrate regeneration solutions formed in waste processing and in deactivation of industrial equipment. It was found that the effects associated with the destruction of the matrix of AXIONIT VPA-2 anion exchanger become more pronounced with the absorbed dose of gamma radiation increasing within the range from 35 to 1000 kGy. A partial destruction of granules of the anion exchanger is observed, its bulk density decreases, swelling in water and nitric acid solution increases, and the plutonium distribution coefficients regularly decrease, but remain rather large. Dynamic tests demonstrated that the irradiation of AXIONIT VPA-2 anion exchanger to an absorbed gamma radiation dose not exceeding 1000 kGy has no significant effect on its sorption properties in extraction of plutonium from nitrate regeneration solutions.

Sequential Extraction of Plutonium from the Bottom Sediments of PA Mayak’s R-4 and R-17 Reservoirs

The speciation of 238, 239, 240Pu in the bottom sediments of two reservoirs of PA Mayak are investigated. The maximum content of plutonium in reservoir R-17 is found in the residual fraction. In reservoir R-4, plutonium is distributed equally in the residual fraction and the fraction bound to organic matter. The fact that plutonium exists in physico-chemically immobile fractions indicate its low biological availability.

Extraction Behaviour of Tris(2-methylbutyl) Phosphate with Fission Products and Heavy Metal Ions

ABSTRACTSystematic screening of various symmetric trialkyl phosphates has led to the conclusion that tris(2-methylbutyl) phosphate (T2MBP), an organophosphate having branched amyl groups, can also be adopted as an alternate extractant to tri-n-butyl phosphate (TBP) for the selective extraction of U(VI) and Pu(IV) during the nuclear fuel reprocessing. In this context, a detailed study on the extraction efficiency of 1.1 M T2MBP/n-dodecane for the extraction of two important nuclear fuel elements, U(VI) and Pu(IV) has been carried out. The cross-current extraction study with U(VI), Pu(IV), and mixed U(VI)–Pu(IV) feed solutions has also been carried out to investigate the extraction and stripping behaviour of the respective metal ions by T2MBP. Results indicated that in contrast to TBP-based systems, there was no occurrence of third-phase formation during the extraction of tetravalent plutonium by T2MBP up to its threshold limit. In addition, its extraction behaviour with certain important fission product elements was also taken up as a part of the study. It has been observed that the extraction of trivalent lanthanides (La, Ce, Nd, Gd), Zr and Tc using T2MBP is marginally lower than is seen for TBP, indicating the possibility of achieving better decontamination factor during reprocessing.

Coordination Structures of Uranium(VI) and Plutonium(IV) in Organic Solutions with Amide Derivatives.

Carbamide and monoamide derivatives are very promising molecules to achieve U(VI) and Pu(IV) extraction and separation from spent nuclear fuels through solvent extraction. Herein, coordination structures of U(VI) and Pu(IV) complexes with carbamide derivatives were characterized using X-ray crystallography as well as infrared, UV-visible, and EXAFS spectroscopies. Coordination structures are compared to those obtained for monoamide derivatives in order to better understand the role of coordination chemistry in extraction properties. Single crystals were first synthesized with a short alkyl chain carbamide analog. Carbamide complexation in the solid state is found analogous to that in the monoamide. In organic solution, upon solvent extraction from nitric acid aqueous solution, it is shown that both amide derivatives can bind in the inner and outer coordination spheres of uranium(VI) and plutonium(IV). The amount of outer sphere coordination complexes increases with the amount of nitric acid. With uranium(VI), at a nitric acid concentration up to 5 mol·L-1, amide derivatives operate predominantly in the inner coordination sphere. In contrast, Pu(IV) coordination geometry is much more sensitive toward acid concentration or ligand structure than U(VI). Pu(IV) changes from inner sphere complexation at 0.5 mol·L-1 HNO3 to mostly outer sphere complexation at 4 mol·L-1. The proportion of outer-sphere complexes is strongly influenced by the ligand structure. Higher Pu(IV) extraction is found to be correlated with the amount of Pu(IV) outer sphere species. Secondary interactions in the outer sphere coordination shell appear to be of primary importance for plutonium extraction.

Unified Framework for Modeling Reactive Extraction of Metals: Illustration on Plutonium(IV) Extraction with Tri-n-butyl Phosphate

Reactive extraction of nuclear elements from aqueous nitric acid solutions using an organic solvent is an important process operation. A thermodynamic model to predict the equilibrium compositions of the two phases will be useful in optimizing the performance of such processes. A general framework is presented in this work for modeling such systems by dividing the complex reactive metal extraction system into appropriate subsystems, whose models are combined to derive the overall model. This approach reduced the number of parameters to be estimated at each stage, making the framework reliable and accurate. The procedure is illustrated for extraction of plutonium (Pu4+) using tri-n-butyl phosphate (TBP) as the organic solvent. This model for Pu extraction correctly accounts for free TBP, the dissociation of nitric acid in concentrated solutions, and formation of hydrolyzed species of plutonium to predict the distribution coefficient of Pu(IV) accurately over the practical range of nitric acid concentration...

Separation of strontium-90 from a highly saline high level liquid waste solution using 4,4′(5′)-[di-tert-butyldicyclohexano]-18-crown-6 + isodecyl alcohol/n-dodecane solvent

Selective and efficient separation of 90Sr from high level waste solution of high salt content will pave the way to manage a problematic waste and also for utilization of recovered 90Sr for societal purpose. In the present work, a process for selective separation of 90Sr is described using 4,4′(5′)-[di-tert-butyldicyclohexano]-18-crown-6 (DtBuCH18C6) dissolved in an industrial diluent system: isodecyl alcohol + n-dodecane. The process solvent is optimized as 0.1 M DtBuCH18C6 + 50% isodecyl alcohol/n-dodecane for satisfactory extraction of strontium at higher nitric acid concentrations (D4-8 M HNO3 = 5–14). Stoichiometry determination by slope analysis method showed 1:1:1 M ratio for Sr2+, DtBuCH18C6 and HNO3 in the extracted complex. Among the inert and active constituents of the waste, extraction of plutonium (IV) (DPu = 0.3) and barium (DBa = 0.7) was observed to some extent, extraction of calcium was almost negligible (DCa = 0.01) while no extraction of americium, cesium, ruthenium, niobium, zirconium and lanthanides was observed. Stripping of the loaded organic was achieved by deionized water, three consecutive contacts each with fresh deionized water results in complete stripping of strontium.

Extraction of plutonium(IV) from acidic feeds using several diamides with a tri-phenyl pyridine centre

Three tri-aryl-pyridine (TAP) diamides containing a pyridine spacer were evaluated for the extraction of Pu(IV) and U(VI) using a process compatible diluent mixture of 95% dodecane + 5% isodecanol. Preferential extraction of Pu(IV) was seen which showed a general increasing trend with increasing aqueous phase nitric acid concentration. The stripping was facile with several reducing/complexing agents and the results of 3 cycles of extraction and stripping were reproducible. The extraction of Am(III), Eu(III), Sr(II) and Cs(I) was < 0.1% suggesting effective decontamination from the fission and activation products which was proven by alpha and gamma ray spectrometric methods.

Extraction of Uranium(VI) and Plutonium(IV) with Tetra-Alkylcarbamides

ABSTRACTThe use of tetra-alkylcarbamides as novel extractants for the separation of uranium(VI) and plutonium(IV) by solvent extraction from spent nuclear fuels is investigated in this study. Batch extraction experiments show that tetra-alkylcarbamides strongly extract U(VI) with high distribution ratios. Plutonium(IV) can be co-extracted with U(VI) at high nitric acid concentration, while high U(VI)/Pu(IV) selectivities can be reached at lower acidity. Loading capacity experiments with high uranium concentrations show that alkyl chains longer than butyl are necessary to avoid third phase formation. Nevertheless, the viscosity of uranium-loaded solvents gets too high with alkyl chains longer than pentyl. Overall, this study shows that with TPU extractant (with four pentyl chains), an efficient co-extraction of uranium and plutonium can be reached (DU,Pu > 1) for a concentration of nitric acid higher than 4 mol•L−1, while the partition between uranium(VI) and plutonium(IV) could be operated even at 2 mol•L−1 nitric acid without redox chemistry.

Effect of chemical environment on the radiation chemistry of N,N-di-(2-ethylhexyl)butyramide (DEHBA) and plutonium retention.

N,N-di-(2-ethylhexyl)butyramide (DEHBA) has been proposed as part of a hydro-reprocessing solvent extraction system for the co-extraction of uranium and plutonium from spent nuclear fuel, owing to its selectivity for hexavalent uranium and tetravalent plutonium. However, there is a critical lack of quantitative understanding regarding the impact of chemical environment on the radiation chemistry of DEHBA, and how this would affect process performance. Here we present a systematic investigation into the radiolytic degradation of DEHBA in a range of n-dodecane solvent system formulations, where we subject DEHBA to gamma irradiation, measure reaction kinetics, ligand integrity, degradation product formation, and investigate solvent system performance through uranium and plutonium extraction and strip distribution ratios. The rate of DEHBA degradation in n-dodecane was found to be slow (G = -0.31 ± 0.02 μmol J-1) but enhanced upon contact with the oxidizing conditions of the investigated solvent systems (organic-only, or in contact with either 0.1 or 3.0 M aqueous nitric acid). Two major degradation products were identified in the organic phase, bis-2-ethylhexylamine (b2EHA) and N-(2-ethylhexyl)butyramide (MEHBA), resulting from the cleavage of C-N bonds, and could account for the total loss of DEHBA up to ∼300 kGy for organic-only conditions. Both b2EHA and MEHBA were also found to be susceptible to radiolytic degradation, having G-values of -0.12 ± 0.01 and -0.08 ± 0.01 μmol J-1, respectively. Solvent extraction studies showed: (i) negligible change in uranium extraction and stripping with increasing absorbed dose; and (ii) plutonium extraction and retention exhibits complex dependencies on absorbed dose and chemical environment. Organic-only conditions afforded enhanced plutonium extraction and retention attributed to b2EHA, while acid contacts inhibited this effect and promoted significant plutonium retention for the highest acidity. Overall it has been demonstrated that chemical environment during irradiation has a significant influence on the extent of DEHBA degradation and plutonium retention.

High effectiveness of pure polydopamine in extraction of uranium and plutonium from groundwater and seawater

Sorption properties of polydopamine (PDA) for uranium and plutonium from an aqueous environment are reported at three different pH values (2, 4 and 6.5–7). In addition to deionized (DI) water, artificial groundwater (GW) and seawater (SW) were used with U uptake close to 100% in each case. PDA polymer has been identified as a material with extremely high sorption capacity Qmax ∼500 mg g−1 of the polymer at pH 6.5 and high selectivity for uranium. Similar high sorption properties are revealed for plutonium uptake. PDA-uranyl and PDA-plutonium interactions responsible for the observed adsorption processes have been addressed with a set of experimental techniques including FTIR spectroscopy, electron microscopy and cyclic voltammetry.

Recovery of uranium and plutonium from pyrochemical salt matrix using supercritical fluid extraction

Pyrochemical reprocessing of spent nuclear fuel involves use of molten salts for the separation of actinides from fission products. Salt waste arising from this process contains residual actinides along with fission products. In the present study, extraction of residual uranium and plutonium from pyrochemical salt matrices was investigated using supercritical fluid extraction (SFE) technique. Supercritical carbon dioxide (SCCO2) containing various ligands was examined to recover actinides from salt matrix with minimum generation of secondary liquid waste. Influence of ligand, salt matrix, temperature, pressure, co-solvent and HNO3 concentration was examined for efficient recovery of actinides. Among the ligands employed in present study, SCCO2 containing trioctylmethylammonium chloride resulted in better recovery of actinides. The recovery of uranium from salt matrix was also demonstrated using semi-preparative scale SFE facility. This study is first of its kind for recovery of actinides from pyrochemical salts using SFE.

A novel approach for extraction of plutonium and americium from aqueous complexing media

A new approach for the extraction of plutonium and americium from aqueous media containing complexing ligands has been investigated using a new extractant N,N′-dioctyl-α-hydroxy acetamide. Studies have been carried out on the extraction of Am(III) and Pu(IV) by this extractant from media containing complexing ligands such as PO43−, C2O42− and F− which are difficult to extract due to the non-extractability of the complexed form of these metal ions. Present study has revealed that, addition of non-complexing anions such as ClO4−, I−, SCN− into the medium containing complexing ligand greatly enhanced extraction of Pu(IV) and Am(III), making the extraction of these nuclides possible.

Selective extraction of plutonium(IV) over uranium(VI), americium(III), europium(III) and zirconium(IV) with bidentate O-phenoxydiamide ligands: experimental and theoretical study

The performance of a series of novel O-phenoxydiamides extracting Zr(IV), Eu(III), U(VI), Pu(IV) and Am(III) was studied. The result of extraction experiments showed that the coordination abilities of these five congeneric O-phenoxydiamides decrease in the order of CycleDODA > PhenylDODA > ButylDODA > HexylDODA > BenzoDODA, the abilities of different metal ions coordinating with the same ligand decrease in the order of Pu(IV) ≫ U(VI) > Zr(IV), Eu(III), Am(III). At HNO3 concentration of 5.0 mol/L, the distribution ratio of BenzoDODA/n-dodecane extracting Pu(IV) is much larger than those of other metal ions, the separation factors of Pu(IV) over Eu(III), Zr(IV), U(VI) and Am(III) are between 20 and 300, which can be used for recovering Pu(IV) selectively from highly acidic liquid waste. The density functional theory (DFT) calculation of the 25 complexes in geometrical structure, Mayer bond order (MBO) and energy was further studied, the result is consistent with the experimental result.

Extraction of uranium and transuranium elements with tert-butylthiacalix[4]arene from carbonate-alkaline solutions

Extraction efficiency of uranium and transuranium elements (Np, Pu, Am and Cm) with tert-butylthiacalix[4]arene TCA from carbonate-alkaline solutions is studied and compared with that of europium (III). Plutonium (III, IV) extraction efficiency with TCA is found to be lower comparing with that of trivalent americium and europium. Extraction efficiency of studied radionuclides decreases as following: Am ≫ Eu ≫ Pu (III), U(VI), Np (V) > Pu (IV) at pH 12. Carbonate concentration increase in aqueous phase suppresses significantly extraction of all studied radionuclides, except americium. This condition can be used for americium individual recovery from complex radioactive carbonate-alkaline solutions.

Review of the Scientific Understanding of Radioactive Waste at the U.S. DOE Hanford Site.

This Critical Review reviews the origin and chemical and rheological complexity of radioactive waste at the U.S. Department of Energy Hanford Site. The waste, stored in underground tanks, was generated via three distinct processes over decades of plutonium extraction operations. Although close records were kept of original waste disposition, tank-to-tank transfers and conditions that impede equilibrium complicate our understanding of the chemistry, phase composition, and rheology of the waste. Tank waste slurries comprise particles and aggregates from nano to micro scales, with varying densities, morphologies, heterogeneous compositions, and complicated responses to flow regimes and process conditions. Further, remnant or changing radiation fields may affect the stability and rheology of the waste. These conditions pose challenges for transport through conduits or pipes to treatment plants for vitrification. Additionally, recalcitrant boehmite degrades glass quality and the high aluminum content must be reduced prior to vitrification for the manufacture of waste glass of acceptable durability. However, caustic leaching indicates that boehmite dissolves much more slowly than predicted given surface normalized rates. Existing empirical models based on ex situ experiments and observations generally only describe material balances and have not effectively predicted process performance. Recent advances in the areas of in situ microscopy, aberration-corrected transmission electron microscopy, theoretical modeling across scales, and experimental methods for probing the physics and chemistry at mineral-fluid and mineral-mineral interfaces are being implemented to build robustly predictive physics-based models.

Innerversusouter sphere metal-monoamide complexation: ramifications for tetravalent &amp; hexavalent actinide selectivity

Extraction of tetravalent and hexavalent plutonium by straight and branched chained monoamides was studied using UV-vis spectroscopy. These results suggest straight monoamides recover outer-sphere species, while branched chain monoamides recover inner-sphere species.

Impact of uranium carbide organics treated by prolonged boiling and electrochemical oxidation upon uranium and plutonium solvent extraction

Abstract The dissolution of uranium or uranium-plutonium carbide fuel in nitric acid leads to ~50% carbon evolved as carbon dioxide, the remainder remains in the solution as soluble organics. These dissolved organic molecules interfere with the solvent extraction of uranium and plutonium by complexing to the actinide ions and decreasing the efficiency of their extraction. Experiments reported here describe two series of experiments assessing the uranium carbide dissolution liquor treatment by prolonged boiling and electrochemical oxidation. Plutonium losses to aqueous and solvent raffinates are observed for untreated liquors, highlighting that mineralisation of dissolved organics is necessary to reduce the complexing effects of organic acids to an extent that permit efficient operation of a solvent extraction process both in the first solvent use (considered here) and for maintaining solvent quality during industrial solvent reuse in the highly active cycle. Solution carbon analysis and 30% TBP solvent extraction batch tests of uranium and plutonium originating from dissolved uranium carbide liquors untreated and after treatment are compared. These experiments demonstrate the reprocessing of uranium carbides by direct dissolution coupled to a mineralisation process, can achieve near quantitative uranium and high plutonium recoveries (99.9%).