Recovery Of Americium Research Articles

Recovery of Americium from Actinide-Lanthanide Fraction of High-Level Waste during Its Transfer to the Oxidation State Am(V)

Recovery of Americium from Actinide-Lanthanide Fraction of High-Level Waste during Its Transfer to the Oxidation State Am(V)

Катионообменное выделение америция при анализе природных вод альфа-спектрометрическим методом

The chromatographic method for isolating americium on a single column with KU-2-8 cation exchange resin from nitric acid solutions is proposed for determination of 241Am in natural waters. The advantages of this simple and rapid procedure are the availability and low cost of equipment, reagents and materials and a small amount of waste. Electrolytic deposition of americium is carried out on stainless steel disks from a solution of ammonium chloride and EDTA-Na2 for 30 minutes. Losses are controlled by the internal standard 243Am. The radiochemical recovery of americium was 60–84% for the model samples, including water with high salinity.

Selective actinide(III) separation using 2,6-bis[1-(propan-1-ol)-1,2,3-triazol-4-yl]pyridine (PyTri-Diol) in the innovative-SANEX process: laboratory scale counter current centrifugal contactor demonstration

Abstract An innovative-SANEX process for the selective separation of the trivalent actinides americium and curium from a simulated PUREX raffinate solution was successfully demonstrated on the laboratory scale using a 16-stage 1 cm annular centrifugal contactor setup. The solvent was composed of 0.2 mol L−1 N,N,N′,N′-tetra-n-octyl-diglycolamide (TODGA) and 5% v/v 1-octanol in a kerosene diluent. Zr(IV) and Pd(II) co-extraction was prevented using trans-1,2-diaminocyclohexane-N,N,N′,N′-tetraacetic acid (CDTA) as a masking agent in the feed. The actinide(III) selective back-extraction was achieved using 2,6-bis[1-(propan-1-ol)-1,2,3-triazol-4-yl]pyridine (PyTri-Diol) in 0.45 mol L−1 HNO3 as a CHON alternative to the sulfur-containing stripping agent used in a previous version of the innovative-SANEX process. The new process described in this paper showed excellent performance for the recovery of An(III). An An(III) product with a quasi-quantitative recovery of americium and curium (≥99.9%) and very good separation from fission and activation products was obtained (decontamination factors ≥4000). Only a slight contamination with Zr and Ru was observed. This test demonstrates the successful use of molecules containing only carbon, hydrogen, oxygen, and nitrogen atoms (so-called CHON molecules) for the selective separation of An(III) from a simulated PUREX raffinate solution. By avoiding sulfur- or phosphorous-containing molecules, the generation of secondary radioactive waste during process operation can be reduced drastically.

Application of Impregnated Resins for the Recovery of Americium and Uranium from Plutonium Contaminated Assorted Analytical Waste

Application of Impregnated Resins for the Recovery of Americium and Uranium from Plutonium Contaminated Assorted Analytical Waste

Batch flowsheet test for a GANEX-type process: the CHALMEX FS-13 process

ABSTRACT 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.

Clean methodology for nuclear laboratory waste remediation: Part-II: Recovery of Americium

As a part of various analytical activities of nuclear laboratories, large volumes of radioactive waste are generated that are extremely hazardous to the environment. The alpha emitting actinides from these wastes must be removed completely (<0.37 Bq/mL) before their discharge for further waste management. In this context, we developed and deployed a novel, simple and clean two step methodology for the recovery of U, Pu and Am from the laboratory waste generated during various chemical quality control activities in nuclear industry. In the first step (reported in Part-I), U and Pu are recovered quantitatively in highly pure form using hollow fibre supported liquid membrane leaving behind all Am and other metallic impurities in the raffinate. In this communication, we report the quantitative recovery of Am from the left-over raffinate using tetraoctyl diglycolamide extraction chromatography. The raffinate after the recovery of actinides (U, Pu and Am) was free from alpha activity that could be managed easily. Additionally, the recovered actinide products were in desired purity for recycling.

All-chromatographic method for the recovery of Americium-241 from solutions of complex composition

This paper reported creating the all-chromatographic technique for the recovery of americium radionuclides from solutions of the complex composition generated in the radiochemical industry. The technique was based entirely on the displacement complexing chromatography; neither additional precipitation nor extraction stages were necessary. The technology used the elution with DTPA and NTA-based solutions containing the citric acid additives and the sequential elution with these eluents, emphasizing the separation of americium from the heavy rare earth elements, lead, and cadmium. The complete separation of americium from lead and cadmium was a peculiar exercise; the elution order of lead and cadmium changed in the presence of the citric acid additives. The recovery and separation of americium from actual radioactive waste solutions of different origins demonstrated 96 and 99% chemical purity of the final product.

Americium Recovery from Highly Active PUREX Raffinate by Solvent Extraction: The EXAm Process. A Review of 10 Years of R&D

ABSTRACT In the framework of the R&D program conducted in France on partitioning and transmutation of minor actinides, a solvent extraction process was developed for separating americium from a PUREX raffinate arising from the dissolution of spent nuclear fuels in nitric acid. The so-called EXAm process uses a mixture of a malonamide (DMDOHEMA, N,N’-dimethyl-N,N’-dioctyl-hexyloxyethyl-malonamide) and HDEHP (di-2-ethylhexylphosphoric acid) in organic phase in combination with a water-soluble complexing agent TEDGA (N,N,N’,N’-tetraethyldiglycolamide) in aqueous phase to enhance Am/Cm separation. Am/Ln separation is then obtained by selective Am stripping with polyaminocarboxylic acids (HEDTA or DTPA) buffered by citric or malonic acid at low acidity. Started in 2008, an important R&D program was conducted at CEA to optimize and model the process. The feasibility of the direct recovery of Am from a genuine PUREX raffinate by solvent extraction was demonstrated in 2010 and reported, for the first time in the literature. R&D was then conducted to adapt the process for a future scale-up and for application from concentrated PUREX raffinates. The results of the final EXAm process performed on a highly active PUREX concentrate are presented after a summary of 10 years of R&D on this process development.

TEHDGA, DMDOHEMA and Mixed Sorbents: Characterization and Am(III) Uptake Properties

The Atalante research facility produces radioactive waste that have to be managed according to specific rules. In particular, solid and liquid wastes need to be transferred to another facility in Marcoule to be evaporated with an alpha activity limit of 16.65 GBq/m3. The radioactive aqueous nitric acid wastes produced in Atalante must be first decontaminated by removing U, Pu and minor actinides and therefore a dedicated decontamination program for radioactive aqueous nitric acid wastes is implemented in Atalante. In particular, the recovery of americium and curium is obtained by extraction chromatography. TEHDGA (N, N, N’, N’-tetra-2-ethylhexyl-3-oxopentane-1,5-diamide), DMDOHEMA ((N,N’-dimethyl-N,N’-dioctylhexyloxyethyl malonamide) and a mixture of both extractants were impregnated on silica gel and studied.Batch experiments were made to determine the performances of such a process. The weight distribution coefficients (Dw) were determined for Am and Nd. Different kinetic behavior was found with both complexants. An effect of nitric acid concentration was observed. For three solids, the americium's uptake increases with acidity. Loading capacities by isotherm and Langmuir isotherm are obtained (between 48mg.g-1 and 94mg.g-1). After determining various parameters in batch extraction experiments, the performances of the solid supports were studied in columns. The breakthrough and elution curves were obtained with neodymium (respectively between 12mg.g-1 and 21mg.g-1; 0.17 and 0.27 L.g-1). Capacity and recovery were optimized with the mixture of DMDOHEMA and TEHDGA.

Extraction Behavior of Americium(III) in Benzoylpyrazolone Dissolved in Pyrrolidinium Based Ionic Liquid

Extraction of Am(III) from nitric acid medium was carried out in a solution of benzoylpyrazolone (HPMBP) dissolved in 1-butyl-1-methylpyrrolidinium bis(trifluoro-methanesulfonyl)imide ([C4mpyr][NTf2]) ionic liquid. The extraction behavior of Am(III) in ionic liquid phase was investigated as a function of various parameters, such as pH of aqueous phase, duration of equilibration, and concentration of ionic liquid and nitrate in aqueous phase. Appreciable extraction was obtained at low pH medium (< pH 2) unlike other diketone based extractant. Extraction of Am(III) in ionic liquid was compared with that observed in a molecular diluent, n-dodecane (n-DD) under similar experimental condition. Remarkably high distribution ratio was observed in IL medium as compared to n-DD medium. The slope analysis of the extraction data was carried out to evaluate the mechanistic aspects of extraction. Extraction mechanism was different from conventional molecular diluent. Recovery of americium from the loaded ionic liquid phase was also studied using nitric acid.

New Insight into the Americium/Curium Separation by Solvent Extraction using Diglycolamides

The liquid–liquid extraction process called EXAm was developed by the CEA to allow the recovery of Americium alone from a PUREX raffinate. Americium is extracted from a highly acidic feed solution (HNO3 4–6 M) by a mixture of two extractants: DMDOHEMA and HDEHP. The Am/Cm selectivity is improved using a specific diglycolamide (TEDGA) as a selective aqueous complexing agent which retains preferentially Cm and heavier lanthanides in the aqueous phase. In this study, the impact of the lipophilicity and steric hindrance of several diglycolamides on the Am/Cm selectivity was investigated in order to understand the enhancement brought by TEDGA. For this purpose, liquid–liquid extraction and partitioning experiments were performed under various conditions.

Simultaneous recovery of plutonium and americium from assorted analytical waste solutions using extraction chromatography

A rapid extraction chromatography based methodology was developed for simultaneous recovery of plutonium and americium from various kinds of analytical waste obtained during chemical quality control of plutonium based nuclear materials using sulphonic acid based actinide™ resin. Efforts were made to understand the effect of initial feed acidity, gamma radiation and the concentrations of Am3+ and Pu4+ on their k d. values. Processing of assorted analytical waste solutions through this method revealed that more than 95 % of Am3+ and 90 % of Pu4+ were adsorbed on the resin while iso- propanol can be successfully employed for the quantitative recovery of both the actinides from the loaded resin phase.

A Novel PVC Based Polymer Inclusion Membrane Containing TODGA as the Extractant for the Pre-concentration of Americium from Acidic Feed Solutions

A series of novel polyvinylchloride (PVC) based polymer inclusion membrane (PIM) containing N,N,N′,N′-tetraoctyl diglycolamide (TODGA) were prepared and applied for recovery of americium from acidic feed solutions. Transport properties of developed PIM for 241Am; were studied in flat sheet supported liquid membrane mode. Quantitative transport (>99.99%) was achieved at four hours with the optimized membrane composition. Optimum PVC PIM was further used for preconcentration of Am from acidic feed solutions. PIM was dipped in required solution with constant stirring. Kinetics of extraction was found to be rapid and about 80% extraction was achieved within two hours. Out of the four strippants used 0.1 M α-HIBA was found to be most suitable and took two hours for complete stripping. Four times volume reduction factors were achieved from acidic waste solutions. Using 24 PVC PIMs, 6.01 mg was successfully trapped from 1.75L waste solution containing 6.2 mg of Am (2.79 M HNO3). The method was quite simple and can be applied for the recovery of Am from laboratory waste.

A Novel Liquid Emulsion Membrane Containing TODGA as the Carrier Extractant for Am Recovery from Acidic Wastes

The recovery of Am has been investigated by Liquid Emulsion Membrane (LEM) containing 0.1 M N,N,N′,N′-tetraoctyl-3-oxapentane diamide (TODGA) + 0.5 M di-n-hexyloctanamide (DHOA) in n-dodecane as extractant. Span 80 (sorbitan monooleate) was used as the surfactant for making the emulsion while 3 M HNO3 and 0.1 M acetate buffer (at pH 4.75) were used as the external and internal phases, respectively. Optimized conditions indicated 2% Span 80 was required for making the emulsions. The effect of different experimental conditions such as varying feed acidity, varying volume ratio, different strippants, and varying TODGA concentration have been studied for the quantitative removal of Am from acidic solution. The higher transport rate has been observed which can be attributed to the higher surface area per unit volume available for mass transport. Demulsification was possible by heating the emulsions to 80°C. Actual acidic radioactive waste has also been tested for the recovery of americium. The results show the applicability of ELM for the removal of Am(III) from waste solutions with a faster rate of transport as compared to those reported previously.

A New Analysis of Composite Materials for Use in Glove-Box Applications

Oil and natural gas companies use 241Am sources for well-logging applications (in the form of americium-beryllium neutron sources). Currently, the domestic supply of 241Am is depleted, and industry is now purchasing sources from Russia. The goal of the Americium Recovery Project (ARP) is to reprocess defense-waste plutonium to recover 241Am that would then be sold to oil and gas companies, providing a safe, secure domestic source for industrial applications. Because the primary radiological concern with an 241Am source is external photon exposure, the radiological workers involved in the ARP will perform operations in glove boxes featuring lead-lined gloves. Given the U.S. mandate for the reduction of lead in industrial settings and the costs associated with the disposal of leaded gloves as mixed waste, alternatives are being considered to the traditional lead-lined gloves used in glove boxes. Several composite materials were previously developed and analyzed for incident photons of energies below 400 keV using the Lambert-Beer law to calculate transmission fractions. This research extends the energy range to 10 MeV and uses a source term of interest to the ARP. Further, the Monte Carlo transport code MCNP5 is used to calculate source-normalized doses using two common response functions: H’(0.07) and H*(10). The results and calculations presented in this research are more detailed than previous calculations and present further rationale for the context-specific selection of a given material.

A review of the demonstration of innovative solvent extraction processes for the recovery of trivalent minor actinides from PUREX raffinate

Abstract The selective partitioning (P) of long-lived minor actinides fromhighly active waste solutions and their transmutation (T) to short-lived or stable isotopes by nuclear reactions will reduce the long-term hazard of the high-level waste and significantly shorten the time needed to ensure their safe confinement in a repository. The present paper summarizes the on-going research activities at Forschungszentrum Jülich (FZJ), Karlsruher Institut für Technologie (KIT) and Institute for Transuranium Elements (ITU) in the field of actinide partitioning using innovative solvent extraction processes. European research over the last few decades, i.e. in the NEWPART, PARTNEW and EUROPART programmes, has resulted in the development of multi-cycle processes for minor actinide partitioning. These multi-cycle processes are based on the co-separation of trivalent actinides and lanthanides (e.g. by the DIAMEX process), followed by the subsequent actinide(III)/lanthanide(III) group separation in the SANEX process. The current direction of research for the development of innovative processes within the recent European ACSEPT project is discussed additionally. This paper is focused on the development of flow-sheets for recovery of americium and curium from highly active waste solutions. The flow-sheets are verified by demonstration processes, in centrifugal contactors, using synthetic or genuine fuel solutions. The feasibility of the processes is also discussed.

Application of hollow fiber supported liquid membrane for the separation of americium from the analytical waste

Americium from analytical solid waste containing U and metallic impurities was separated using hollow fiber supported liquid membrane (HFSLM) technique impregnated with DHOA–TODGA from nitric acid medium. An aliquot of 5 g of the solid waste containing Am (19.95 mg) as minor actinide and of U (2,588 mg), Fe (1,360 mg), Ca (1,810 mg) and Na (3,130 mg) as major impurities was processed. The feed solution obtained after the dissolution of the residue in ~4 M HNO3 was passed through HFSLM module. In the first stage using 1 M DHOA–dodecane U was recovered while Am and other impurities were left in the raffinate. In the second stage, 0.5 M DHOA + 0.1 M TODGA/dodecane was used for the separation of Am from other impurities. Though, majority of the elements were separated in this cycle, Ca was co extracted along with the americium. CMPO extraction chromatographic technique was used for further separation of americium from Ca. Significant decontamination factors were achieved in this three step separation process with respect to U, Fe, Na and Ca with ~77 % recovery of americium.

A highly efficient solvent system containing functionalized diglycolamides and an ionic liquid for americium recovery from radioactive wastes

Three room temperature ionic liquids (RTILs), viz. C(4)mim(+)·PF(6)(-), C(6)mim(+)·PF(6)(-) and C(8)mim(+)·PF(6)(-), were evaluated as diluents for the extraction of Am(III) by N,N,N',N'-tetraoctyl diglycolamide (TODGA). At 3 M HNO(3), the D(Am)-values by 0.01 M TODGA were found to be 102, 34 and 74 for C(4)mim(+)·PF(6)(-), C(6)mim(+)·PF(6)(-) and C(8)mim(+)·PF(6)(-), respectively. The extraction of Am(III) decreased with increasing feed acidity for all three diluents, indicating an ion exchange mechanism for the extraction. The stoichiometry of the extracted species suggested that two TODGA molecules were associated with Am(III) during the extraction for all three RTILs and the conditional extraction constants have been determined. The D(M)-values for different metal ions followed the order: 75 (Am(III)) > 30.7 (Pu(IV)) > 3.9 (Np(IV)) > 1.19 (Pu(VI)) > 0.52 (U(VI)) > 0.12 (Cs(I)) > 0.024 (Sr(II)). The distribution behaviour of Am(III) was also studied with a recently synthesized calix[4]arene-4DGA (C4DGA) extractant dissolved in C(8)mim(+)·PF(6)(-). Using this extractant diluent combination, the D(Am)-value was 194 at 3 M HNO(3) using 5 × 10(-5) M C4DGA, suggesting a very high distribution coefficient at very low extractant concentrations. The stoichiometry of the extracted species containing Am was found to be 1:2 (M:L) in C(8)mim(+)·PF(6)(-). The thermodynamics of the extraction was also studied for both extractants in C(8)mim(+)·PF(6)(-). The use of RTILs gives rise to significantly improved extraction properties than the commonly used n-dodecane and an unusual increase in separation factor values was seen for the first time which can lead to selective separation of Am from wastes containing a mixture of U, Pu and Am.

Solid-phase extraction of plutonium(IV) an americium(III) using N-benzoylphenylhydroxylamine and its derivatives

The recovery and separation of plutonium(IV) and americium(III) by solid-phase extraction (SPE) on alkylated silica gel S16 modified with N-benzoylphenylhydroxylamine (BPHA) and with its derivatives was studied. BPHA was modified by introducing into the p-position of the phenyl ring of electronactive substituents that differ in their hydrophobicity: CH3, Ph, Cl, F, and NO2. The SPE of plutonium(IV) and americium(III) was studied in the range of acidities from pH 8 to 1 M HNO3. The recovery and separation of these elements was shown to depend on the nature of the substituent, aqueous acidity, and the preparation of S16 to SPE experiments.

Recovery of americium from analytical solid waste containing large amounts of uranium, plutonium and silver

Trace metallic impurity analysis by spectroscopic techniques is one of the important steps of chemical quality control of nuclear fuel materials. Depending on the burn-up and the storage time of the fuel, there is an accumulation of 241Am in plutonium based fuel materials due to β decay of 241Pu. In this paper, attempts were made to develop a method for separation of 241Am from 1.2 kg of analytical solid waste containing 70% U, 23% Pu, 5% Ag and 1–2% C as major constituents along with other minor constituents generated during trace metal assay of plutonium based fuel samples by d. c. arc carrier distillation atomic emission spectrometry. A combination of ion exchange, solvent extraction and precipitation methods were carried out to separate ~45 mg of 241Am as Am(NO3)3 from 15 L of the analytical waste solution. Dowex 1×4 ion exchange chromatographic method was used for separation of Pu whereas 30% TBP–kerosene was utilized for separation of U. Am was separated from other impurities by fluoride precipitation followed by conversion to nitrate. The recovery of Pu from ion exchange chromatographic separation step was ~93% while the cumulative recovery of Am after separation process was found to be ~90%.