Recovery Of Actinides Research Articles

Demonstration of the Separation of Actinides from Lanthanides By Electrolysis with in-Situ Voltammetry

Electrolysis can be used to recover actinides and lanthanides from electrorefiner salt to facilitate fission product removal and salt recycle in pyrochemical flowsheets. High separability between actinides and lanthanides is a key goal in meeting pyrochemical flowsheet requirements such as maximizing actinide recovery and minimizing salt discharge to waste. The core separation unit in a pyrochemical system is the electrorefiner that separates uranium and transuranics (TRU) from a metallic fuel or reduced oxide fuel through oxidation of the metals at an anode and electrodeposition of the metal at a cathode. After this separation, in which uranium and U+TRU are recovered by cathodic deposition, the electrorefiner salt consists of a mixture of dissolved actinide, and lanthanide chlorides and Group 1 and 2 fission product chlorides in eutectic LiCl/KCl. The ratio of actinides and lanthanides are of critical importance to the purity of the actinide product recovered and must be carefully controlled to minimize lanthanide content in the recovered U+TRU product. Therefore periodic drawdown of the accumulated lanthanide chlorides is required. Reduction of the lanthanide chlorides to metals offers the greatest degree of actinide/lanthanide separation. Because the lanthanide chlorides are more stable than actinide chlorides, it is necessary to first drawdown all of the actinide species. Once the actinides are removed, a similar drawdown operating at more reducing conditions can be used to remove lanthanides from the molten salt. Although the actinide free salt could be disposed of in a suitable wasteform, a lanthanide drawdown followed by concentration of Group 1 and 2 fission products allows for recycle of cleaned electrorefiner salt, thus minimizing the volume of high-level waste generated by the overall process. Drawdown of actinides and lanthanides by electrolysis has two advantages over the addition of a chemical reductant. First, addition of a chemical reductant adds to the waste volume. Secondly, electrolysis simplifies recovery of the actinides and lanthanides because reduction of the metals occurs in the localized area of the cathode surface. Thus the goal of this work is to examine the degree of group separability that can be achieved between actinides and lanthanides in an electrolysis drawdown operation. In a set of electrolysis feasibility experiments, deposition of metals occurs at a central graphite cathode and chlorine gas evolves at three graphite anodes placed equidistant from the cathode. The desired depositing metals are selected by careful control of electrode surface area and potential versus a Ag/AgCl reference electrode. An electrolysis cell of novel design was paired with a chlorine gas scrubbing system to perform the experiments. A set of tungsten working and counter electrodes were used to conduct in-situ voltammetry of the bulk. Voltammetry was used to monitor the concentration of actinides and lanthanides present in the bulk salt during electrolysis. The feasibility demonstration was conducted in three phases, first the reduction of uranium from eutectic LiCl/KCl containing UCl3, second the reduction of uranium from the same salt with the addition of GdCl3, and finally the separation of a group of actinide chlorides from a group of lanthanide chlorides in the same salt. The uranium drawdown experiments successfully removed uranium from the salt and generated chlorine gas. Analytical results from filtered salt samples verified a reduction in UCl3 concentration corroborating the change observed by in-situ voltammetry. Comparison with known starting concentrations allows for the calculation of actinide concentrations from voltammetry data. Products from drawdown experiments had nodule like morphology and heavy salt entrainment, which complicated product recovery. Although the initial drawdown experiments achieved consistently low yields with current efficiencies estimated at below 20%, sequential drawdown of the actinides, an actinide – lanthanide mixture and finally lanthanides was achieved. The lower current efficiency may have been caused by design limitations in the test system, back reactions with higher valence uranium ions in the salt, and/or loss of product not firmly adhered to the cathode. Re-chlorination by dissolved chlorine was eliminated as a possible cause due to low solubility of chlorine in the salt. Future testing of electrolysis for actinide - lanthanide separations will focus on improving product recovery, process yield and efficiency. Government License Notice The submitted manuscript was created by UChicago Argonne, LLC, Operator of Argonne National Laboratory (“Argonne”). Argonne, a U.S. Department of Energy Office of Science laboratory, is operated under Contract No. DE-AC02-06CH11357. The U.S. Government retains for itself, and others acting on its behalf, a paid-up nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government.

Investigations on synthesis, coordination behavior and actinide recovery of unexplored dicyclohexylphosphinic acid

Dicyclohexylphosphinic acid (DcyHPA) and some of its metal complexes with lanthanides (La, Ce, Gd, Sm and Nd) and actinides (U and Th) were prepared and subsequently the structural composition was elucidated. The single crystal XRD data and theoretical analysis revealed that the ligand (DcyHPA) exists as a dimer in solid state. DcyHPA showed dual nature for the extraction of U(VI), Th(IV) and Am(III) from nitric acid medium i.e. at lower acidity, the ligand behaves as an acidic extractant and extracted the metal through cation exchange mechanism while solvation mechanism played a major role for extraction at higher acidity. Additionally, density functional theory (DFT) calculations were demonstrated to understand the electronic structure of thorium-DcyHPA metal complex formed during the extraction process.

Dissolution of WWER-1000 spent nuclear fuel in a weakly acidic solution of iron nitrate and recovery of actinides and rare earth elements with TBP solutions

It is demonstrated on real solutions of samples of spent nuclear fuel (SNF) from WWER-1000 reactors (1000-MWel water-cooled water-moderated energy reactors) that weakly acidic solutions of iron(III) nitrate at the molar ratio Fe(III): U ≥ 2.0 dissolve SNF with quantitative transfer of U and Pu into the solution. In the process, Fe partially precipitates in the form of a basic salt precipitate together with a part of the fission products (>90% of Ru, ~90% of Мо, >60% of Tc, and 40% of Zr) already in the step of the fuel dissolution. Cs, Eu, and Am pass into the solution together with U and Pu. With the required conditions followed, U and Pu can be separated from the solution by precipitation of their peroxides or quantitatively extracted from this solution with 30% TBP in Isopar L. The presence of ≥1 M Fe(NO3)3 in the solution considerably increases the distribution ratios of TPE and REE, which allows their recovery from a weakly acidic nitrate solution to be also performed with 30% TBP in a diluent. This process can serve in the future as a basis for the development of a new integrated technology combining the PUREX process with TPE partitioning using a common extractant.

Multi-podant diglycolamides and room temperature ionic liquid impregnated resins: An excellent combination for extraction chromatography of actinides

Extraction chromatography resins, prepared by impregnating two multi-podant diglycolamide ligands, viz. diglycolamide-functionalized calix[4]arene (C4DGA) and tripodal diglycolamide (T-DGA) dissolved in the room temperature ionic liquid 1-butyl-3-methylimidazolium bis(trifluoromethanesulfonyl)amide (RTIL: C4mimTf2N) on Chromosorb-W (an inert solid support), gave excellent results for the removal of trivalent actinides from acidic waste solutions. Distribution coefficient measurements on several metal ions showed selective sorption of Am(III) over hexavalent uranyl ions and other fission product elements such as strontium and cesium. The sorbed metal ions could be efficiently desorbed with a complexing solution containing guanidine carbonate and EDTA buffer. The sorption of Am(III) on both resins followed pseudo-second order rate kinetics with rate constants of 1.37×10−6 and 6.88×10−7g/cpmmin for T-DGA and C4DGA resins, respectively. The metal sorption on both resins indicated the Langmuir monolayer chemisorption phenomenon with Eu(III) sorption capacities of 4.83±0.21 and 0.52±0.05mg per g of T-DGA and C4DGA resins, respectively. The results of column studies show that these resins are of interest for a possible application for the recovery of hazardous trivalent actinides from dilute aqueous solutions.

Assessment of monoamide extractants and solid supports as new extraction chromatographic materials

Abstract The potential of monoamide-based extraction chromatographic materials for actinide recovery was assessed by monitoring uranium partitioning into six different materials. These materials were comprised of two solid supports (Amberlite XAD 4 and XAD 7) coated with one of three extractants: di-2-ethylhexyl butyramide (DEHBA), di-2-ethyhexyl isobutryamide (DEHiBA), or di-2-ethyhexyl acetylamide (DEHAA). This report considers differences in uranium partitioning depending on the solid support, extractant, aqueous phase conditions and amount of extractant coating on a given support. Studies indicate the acetyl amide extractant coated onto XAD 7 is a top candidate for further investigation. This is an interesting finding as the acetyl amide is not generally considered a particularly useful extractant for solvent extraction separations due to its tendency to form a third phase. This information contrasts the general paradigm that a given extractant can be equally useful for extraction chromatographic or solvent extraction separations and indicates that the development of future classes of extraction chromatographic materials should carefully consider the role of alkyl groups in their optimization. Additionally, the efficacy of the developed monoamide materials are compared with more classically utilized organophosphorous materials to consider opportunities for advancement in f-element separations.

Zirconium behaviour during electrorefining of actinide-zirconium alloy in molten LiCl-KCl on aluminium cathodes

A pyrochemical electrorefining process for the recovery of actinides from metallic nuclear fuel based on actinide-zirconium alloys (An–Zr) in a molten salt is being investigated. In this process actinides are group-selectively recovered on solid aluminium cathodes as An–Al alloys using a LiCl–KCl eutectic melt at a temperature of 450 °C. In the present study the electrochemical behaviour of zirconium during electrorefining was investigated. The maximum amount of actinides that can be oxidised without anodic co-dissolution of zirconium was determined at a selected constant cathodic current density. The experiment consisted of three steps to assess the different stages of the electrorefining process, each of which employing a fresh aluminium cathode. The results indicate that almost a complete dissolution of the actinides without co-dissolution of zirconium is possible under the applied experimental conditions.

Development and Testing of Diglycolamide Functionalized Mesoporous Silica for Sorption of Trivalent Actinides and Lanthanides.

Sequestration of trivalent actinides and lanthanides present in used nuclear fuel and legacy wastes is necessary for appropriate long-term stewardship of these metals, particularly to prevent their release into the environment. Organically modified mesoporous silica is an efficient material for recovery and potential subsequent separation of actinides and lanthanides because of its high surface area, tunable ligand selection, and chemically robust substrate. We have synthesized the first novel hybrid material composed of SBA-15 type mesoporous silica functionalized with diglycolamide ligands (DGA-SBA). Because of the high surface area substrate, the DGA-SBA was found to have the highest Eu capacity reported so far in the literature of all DGA solid-phase extractants. The sorption behavior of europium and americium on DGA-SBA in nitric and hydrochloric acid media was tested in batch contact experiments. DGA-SBA was found to have high sorption of Am and Eu in pH 1, 1 M, and 3 M nitric and hydrochloric acid concentrations, which makes it promising for sequestration of these metals from used nuclear fuel or legacy waste. The kinetics of Eu sorption were found to be two times slower than that for Am in 1 M HNO3. Additionally, the short-term susceptibility of DGA-SBA to degradation in the presence of acid was probed using (29)Si and (13)C solid-state NMR spectroscopy. The material was found to be relatively stable under these conditions, with the ligand remaining intact after 24 h of contact with 1 M HNO3, an important consideration in use of the DGA-SBA as an extractant from acidic media.

Actinides recovery from irradiated metallic fuel in LiCl–KCl melts

Electrorefining of irradiated metallic fuels was successfully demonstrated: Actinides (U, Pu, Np, Am and Cm) in the fuels were dissolved in LiCl–KCl melts with high dissolution ratios, while U was selectively deposited on a solid cathode and the simultaneous recovery of actinides in a liquid Cd cathode was confirmed. The behavior of actinides, the fuel matrix stabilizer Zr and fission products such as lanthanide, alkaline, alkaline earth and noble metal, at the electrorefining is discussed based on the ICP-MS analysis of the samples taken from molten salt electrolyte, anode fuel residues and cathode deposits.

Hydroxyapatite Biosynthesis by a Serratia sp. and Application of Nanoscale Bio-HA in the Recovery of Strontium and Europium

ABSTRACTA Serratia sp. expresses a high level of acid phosphatase when grown continuously under carbon limitation. In the presence of CaCl2, biosynthesis of nanocrystalline hydroxyapatite (bio-HA) was achieved by utilizing phosphate released via enzymatic cleavage of an applied substrate (glycerol 2-phosphate: G2P). Hydroxyapatite crystals were identified by energy dispersive X-ray emission (EDX) and selected area diffraction (SAD). X-ray powder diffraction (XRD) analysis gave a mean crystallite size of ∼21–32 nm, with the smallest crystals (21–24 nm) obtained using 1 mM Ca2+ and 1 mM G2P. The uptake of Eu3+ and Sr2+ by bio-HA made by continuously pregrown cells (0.42 mg/mg and 0.043 mg/mg respectively) was ∼20% greater for Sr2+ than was previously reported for bio-HA material of size ∼40 nm made by batch-pregrown cells, while the corresponding uptake of Eu3+ was increased by > 1.8-fold. This was attributed to the localization of Eu (III) at grain boundaries by reference to previous work and highlights the potential of bio-HA as a sequestration agent for recovery of rare earth elements and trivalent actinides.

Complete Recovery of Actinides from UREX-like Raffinates Using a Combination of Hard and Soft Donor Ligands. II. Soft Donor Structure Variation

The feasibility of simultaneous separation of uranium, neptunium, plutonium, americium, and curium from a simulated dissolved used fuel simulant adjusted to 1.0 M nitric acid is investigated using a mixture of the soft donor bis(bis-3,5-trifluoromethyl)phenyl) dithiophosphinic acid (“0”) and the hard donor synergist trioctylphosphine oxide (TOPO) dissolved in toluene. The results reported in this work are compared to our recent demonstration of a complete actinide recovery from a simulated dissolved fuel solution using a synergistic combination of bis(o-trifluoromethylphenyl)dithiophosphinic acid (“1”) and TOPO dissolved in either toluene or trifluoromethylphenyl sulfone. While the extraction efficiency of americium was enhanced for the liquid-liquid system containing “0”, enabling to accomplish a trivalent An/Ln separation at 1.0 M HNO3, the extraction of neptunium was drastically diminished, relative to “1”. The partitioning behavior of curium was also negatively impacted, introducing an effective opportunity for americium/curium separation. Radiometric and spectrophotometric studies demonstrate that the complete actinide recovery using the solvent based upon “0” and TOPO is not feasible. In addition, the importance of radiolytic degradation processes is discussed through the comparisons of extraction properties of liquid-liquid systems based on both soft donor reagents.

A model study on multistep recovery of actinides based on the difference in diffusion coefficients within liquid metal

Abstract This study presents an effective method for additional recovery of residual actinides in liquid electrodes after the electrowinning process of pyroprocessing. The major distinctive feature of this method is a reactor with multiple reaction cells separated by partition walls in order to improve the recovery yield, thereby using the interelement difference in diffusion coefficients within the liquid electrode and controlling the selectivity and purity of element recovery. Through an example of numerical simulation of the diffusion scenarios of individual elements, we verified that the proposed method could effectively separate the actinides (U and Pu) and rare-earth elements contained in liquid cadmium. We performed a five-step consecutive recovery process using a simplified conceptual reaction cell and recovered 58% of the initial amount of actinides (U + Pu) in high purity (≥ 99%).

Pyrochemical extraction analysis of an immiscible molten LiCl–KCl/Cd system

A pyrochemical extraction is one of the promising unit operations used in the pyrometallurgical process for the separation of actinides from fission products mixture with the aid of extracting the solvent and oxidant. A rigorous computational model for the simulation of multi-stage counter-current oxidative extraction was developed by using model equations, composed of a material balance and distribution relationships between the molten-salt and metal phase stream. The solutions were determined using the equation tearing algorithm. It was observed that this modeling approach allows us to provide simulations for the effect of the operating variables on the recovery of group actinides.

Analysis of uranium and plutonium following adsorption on actinide resin

A method has recently been developed at AWE for the pre-concentration, recovery, separation and measurement of actinides using Eichrom’s Actinide™ Resin. The method was specifically developed to enable successful isotopic analysis of uranium and plutonium at trace environmental levels using isotope dilution thermal ionisation mass spectrometry. This paper summarises the development of the method, associated mass spectrometry results and the challenges which were overcome in recovering and purifying uranium and plutonium spikes (233U and 239Pu) during method development. Actinide™ Resin is known to be very good at selectively pre-concentrating actinides from slightly acidic and neutral solutions making this method attractive for measuring uranium and plutonium isotopic ratios in environmental samples.

Evaluation study on a macroporous silica-based isohexyl-BTP adsorbent for minor actinides separation from nitric acid medium

Abstract To separate the long-lived minor actinides (MA = Am, Cm) from high level liquid waste (HLLW), we have been studying a partitioning technology named “MAREC” (Minor Actinides Recovery from HLLW by Extraction Chromatography) process which consists of two separation columns packed with silica-based adsorbents. In this work, we prepared the silica-based 2,6-bis(5,6-diisohexyl-1,2,4-triazin-3-yl)pyridine (isohexyl-BTP/SiO2-P) adsorbent used in the second column (0.01 ∼ 0.1 mol dm –3 HNO3) in the MAREC process for the separation of trivalent minor actinides (MA(III)) from trivalent rare earth elements (RE(III)) which are contained in HLLW. The content of extracting agent isohexyl-BTP in isohexyl-BTP/SiO2-P was as high as 33.3% of the total mass of the adsorbent. The adsorbent exhibited high affinity and selectivity for Am(III) over RE(III) not only in nitric acid solution but also in nitrate solution. Investigation showed that only a small part of isohexyl-BTP molecules inside isohexyl-BTP/SiO2-P leaked into 0.01 mol dm –3 HNO3 with or without γ-irradiation. The adsorption performance of isohexyl-BTP/SiO2-P did not evidently change after 120 d of contact with 0.01 mol dm –3 nitric acid. Furthermore, after γ-irradiated in 0.01 mol dm –3 HNO3 under the experimental conditions, isohexyl-BTP/SiO2-P exhibited higher adsorption for RE(III) than before γ-irradiation.

Sorption recovery of U(VI), Pu(IV), and Am(III) from nitric acid solutions with solid-phase extractants based on Taunit carbon nanotubes and polystyrene supports

Sorption procedures were developed for recovering U(VI), Pu(IV), and Am(III) with solid-phase extractants (SPEs) prepared by impregnation of Taunit carbon nanotubes and polystyrene supports with diphenyl( dibutylcarbamoylmethyl)phosphine oxide (CMPO) and tri-n-octylphosphine oxide (TOPO) The impregnation and actinide recovery were performed in the batch mode and using microcolumns. Procedures for support impregnation and SPE preparation are described. Conditions were found for sorption recovery of U(VI), Pu(IV), and Am(III) from 3 M HNO3 solutions. The possibility of actinide desorption was demonstrated. The effect of macrocomponents on the degree of actinide recovery was examined.

Recovery of actinides from actinide–aluminium alloys by chlorination: Part II

A chlorination route is being investigated for recovery of actinides from actinide–aluminium alloys, which originate from pyrochemical recovery of actinides from spent metallic nuclear fuel by electrochemical methods in molten LiCl–KCl. In the present work, the most important steps of this route were experimentally tested using U–Pu–Al alloy prepared by electrodeposition of U and Pu on solid aluminium plate electrodes. The investigated processes were vacuum distillation for removal of the salt adhered on the electrode, chlorination of the alloy by chlorine gas and sublimation of the AlCl3 formed. The processes parameters were set on the base of a previous thermochemical study and an experimental work using pure UAl3 alloy. The present experimental results indicated high efficiency of salt distillation and chlorination steps, while the sublimation step should be further optimised.

Development and demonstration of innovative partitioning processes (i-SANEX and 1-cycle SANEX) for actinide partitioning

Two full continuous countercurrent processes for the recovery of trivalent actinides (An(III)) from a simulated PUREX raffinate solution (HAR) were developed and demonstrated using miniature laboratory-scale centrifugal contactors: the innovative SANEX and the 1-cycle SANEX processes. The innovative SANEX process was successfully demonstrated in a 16 + 16 stage flow sheet. In the first 16 stages, Am(III), Cm(III), and Ln(III) were quantitatively co-extracted from the HAR by a TODGA-based extractant. The decontamination factors (DFs) for the major non-lanthanide fission products were >103. The loaded extractant was then subjected to two stripping steps using the second 16 stages of the flow sheet. The first stripping step concerned the selective stripping of Am(III) + Cm(III) at fairly high acidity (0.35 ML−1 HNO3) with the new hydrophilic N-chelating selective ligand SO3-Ph-BTP. High An(III) recoveries >99.8% were achieved with high decontamination factors toward the trivalent lanthanides. In the second step, the residual stripping of Ln(III) from loaded organic phase was carried out quantitatively using 0.5 ML−1 citric acid solution at pH 3. However, Ru proved to be the only exception and remained to a large extent (12.8%) in the spent extractant. A more challenging route also studied at our laboratories is the 1-cycle SANEX process, i.e. direct An(III) separation from HAR using an extractant mixture of CyMe4BTBP and TODGA in 1-octanol/TPH diluent. A demonstration process was also successfully implemented using a 16 + 16 stage flow sheet on the above-mentioned laboratory-scale centrifugal contactor rig. It was demonstrated that a selective extraction and high recovery of >99.8% of Am(III) and >99.4% Cm(III) was achieved with low contamination of fission products. Both new processes are major contributions to the field of partitioning and important steps forward toward the industrial implementation of MA partitioning.

Progress towards the Full Recovery of Neptunium in an Advanced PUREX Process

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.

Complete Recovery of Actinides from UREX-like Raffinates using a Combination of Hard and Soft Donor Ligands

Studies have shown a much enhanced differentiation between trivalent actinides and trivalent lanthanides when substituted alkyl dithiophosphinic acids are utilized as the active phase transfer reagents. We present a study evaluating the use of a synergistic combination of bis(o-trifluoromethylphenyl)dithiophosphinic acid and trioctylphosphine oxide for the complete actinide recovery from UREX-like raffinates. A complete single-step separation of uranium, neptunium, plutonium, and americium from aqueous mixtures containing 0.5 M nitric acid and a total fission product content of 5.5 g/L (45 mM) has been demonstrated.

Recovery of plutonium from polymeric waste matrices using supercritical fluid extraction

Polymers are widely used materials and are indispensable in handling radioactive materials in fumehoods, gloveboxes, etc. Supercritical fluid extraction (SFE) technique is widely employed for the removal of radioactive contaminants from various matrices using supercritical carbon dioxide (Sc-CO2) modified with suitable ligands as an extraction medium. In the present study, recovery of plutonium from various polymeric matrices, such as neoprene, polyvinyl chloride (PVC) and surgical gloves has been demonstrated using Sc-CO2 modified with n-octylphenyl N,N-diisobutyl carbamoylmethylphosphine oxide (ϕCMPO) in methanol. The matrix was initially subjected to neat Sc-CO2 extraction to prevent co-extraction of organic compounds during SFE of plutonium; in the initial extraction with Sc-CO2, organic compounds added as additives to the polymeric matrix were extracted. In the initial studies, SFE method was developed for the complete recovery of plutonium from simulated polymeric waste matrices. Subsequently, the actual waste, i.e. plutonium present in PVC matrix was processed from a 0.1 L extraction vessel for its recovery. The plutonium recovery was found to be ∼97% from the actual PVC waste matrix. The technique was also demonstrated for the recovery of important actinides e.g. americium from simulated neoprene waste matrix. As the disposal of polymer waste is of major concern in nuclear waste management, the SFE technique developed in the present study offers a potential opportunity for the recovery of actinides with generation of minimum secondary waste.