Characterization of Portuguese geomaterials,the clay component of <Emphasis Type="Italic">rañas</Emphasis>, as potential liners for low and intermediate radioactive disposal sites

This paper describes the development of a separation method for americium from the effluents emanating from anion exchange column, used for the recovery of plutonium from analytical waste solutions. The waste contained uranium, sodium, calcium and iron as the major impurities as estimated by ICP-AES method. ~99% pure americium was obtained by three separation steps using solvent extraction and extraction chromatography techniques. In the first step, uranium was quantitatively separated by giving five contacts of equal volumes of 30% TBP in n-dodecane. Fe and Na were separated in the next step using 0.1 M TODGA + 0.5 M DHOA as the extractant. In the last step, Am was separated from the co-extracted Ca (about 76%) using CMPO loaded extraction chromatographic column. The overall recovery was >80% with decontamination factor (D.F.) from the impurities being >3000 while the purity of the product was 99%.     

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.     

Extraction of actinides and fission products by octyl(phenyl)-N,N-diisobutylcarbamoylmethyl-phosphine oxide from nitric acid media

Extraction of promethium(III), uranium(VI), plutonium(IV), americium(III), zirconium(IV), ruthenium(III), iron(III) and palladium(II) has been carried out with a mixture of octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tributyl phosphate (TBP) in dodecane. The effects of nitric acid, TBP and CMPO concentrations on the extraction of these metal ions have been studied. The nature of the species of the above metal ions extracted into the organic phase has been suggested.     

Purification of americium from assorted analytical waste in hydrochloric acid medium

During the simultaneous extraction of plutonium and uranium using anion exchange chromatographic technique from analytical waste in hydrochloric acid medium, ²⁴¹Am which is invariably present in the plutonium bearing fuel samples remains in the effluent. A two step separation scheme was developed for the recovery and purification of Am from the assorted waste to facilitate the disposal of large volume of aqueous waste and the purified Am solution was utilized for spectroscopic investigation. The separation scheme involved solvent extraction using 0.1 M TODGA + 0.5 M DHOA for separation of americium from Fe, Pb, Ni and Na followed by extraction chromatographic technique using CMPO on inert support as stationary phase for separation of Ca from Am. A systematic study on the extraction behavior of Am from hydrochloric acid medium revealed that out of four extraction systems well known for actinide partitioning namely 0.1 M TODGA + 0.5 M DHOA, 1 M DMDBTDMA, 0.2 M CMPO + 1.2 M TBP and 30% TRPO, only 0.1 M TODGA + 0.5 M DHOA extracts americium from 7.5 M HCl feed acidity. A comparative study involving CMPO solvent extraction and column chromatographic technique revealed that elution of Am from column is satisfactory as compared to inefficient stripping of Am from organic phase in solvent extraction technique using 0.1 M HNO₃. The purity of the final solution was checked for 17 elements of interest and was found to be 98% pure, while the overall recovery of this two step separation scheme was found to be 95%.     

Extraction of americium(VI) by a neutral phosphonate ligand

Recently the use of the more unusual hexavalent oxidation state of americium has been receiving increased attention for the purpose of developing an efficient Am/Cm or Am/lanthanide separation system. We have already demonstrated the feasibility of performing this separation with 30% TBP in dodecane, and are now looking at different extractants to increase Am(VI) distribution ratios. Following on from this the extraction of bismuth oxidized americium from nitric acid solutions by dibutyl butyl phosphonate has been studied. The results of this study indicate that increasing the basicity of the extractant molecule has significantly improved the extraction efficiency.     

Separation and Recovery of Uranium and Plutonium from Oxalate Supernatant

The separation of uranium and plutonium from oxalate supernatant, obtained after precipitating plutonium oxalate, containing ~10 g/l uranium and 30–100 mg/l plutonium in 3M HNO₃ and 0.10–0.18M oxalic acid solution has been carried out. In one extraction step with 30% TBP in dodecane: ~92% of uranium and ~7% of Pu is extracted. The raffinate containing the remaining U and Pu is extracted with 0.2M CMPO+1.2 M TBP in dodecane and near complete extraction of both the metal ions is achieved. The metal ions are back extracted from organic phases using suitable stripping agents. The recovery of both the metal ions separately is >99%. The uranium species extracted into the TBP phase from the HNO₃+oxalic acid medium was identified as UO₂(NO₃)₂·2TBP.     

Tributylphosphate extraction behavior of bismuthate-oxidized americium

Higher oxidation states of americium have long been known; however, options for their preparation in acidic solution are limited. The conventional choice, silver-catalyzed peroxydisulfate, is not useful at nitric acid concentrations above about 0.3 M. We investigated the use of sodium bismuthate as an oxidant for Am (3+) in acidic solution. Room-temperature oxidation produced AmO 2 (2+) quantitatively, whereas oxidation at 80 degrees C produced AmO 2 (+) quantitatively. The efficacy of the method for the production of oxidized americium was verified by fluoride precipitation and by spectroscopic absorbance measurements. We performed absorbance measurements using a conventional 1 cm cell for high americium concentrations and a 100 cm liquid waveguide capillary cell for low americium concentrations. Extinction coefficients for the absorbance of Am (3+) at 503 nm, AmO 2 (+) at 514 nm, and AmO 2 (2+) at 666 nm in 0.1 M nitric acid are reported. We also performed solvent extraction experiments with the hexavalent americium using the common actinide extraction ligand tributyl phosphate (TBP) for comparison to the other hexavalent actinides. Contact with 30% tributyl phosphate in dodecane reduced americium; it was nevertheless extracted using short contact times. The TBP extraction of AmO 2 (2+) over a range of nitric acid concentrations is shown for the first time and was found to be analogous to that of uranyl, neptunyl, and plutonyl ions.     

Characterization of hot particles in surface soil around the Chernobyl NPP

An ICP-AES method for the analysis of trace amounts of lanthanides and yttrium in sodium or magnesium diuranate samples (yellow cake) and other beneficiation product generated during the uranium extraction process (hydrometallurgy) from its ores is described. Most of the matrix elements are removed by an initial oxalate precipitation of lanthanides using calcium as carrier. A solvent extraction procedure using a mixture of mono 2-ethylhexyl dihydrogen phosphate (H₂MEHP) and bis (2-ethylhexyl) hydrogen phosphate (HDEHP) is used for the removal of calcium, iron and the occluded uranium. A combination of oxalate precipitation and solvent extraction procedure is applied for the selective separation and preconcentration of traces of lanthanides from yellow cake and iron cake samples. The solvent extraction procedure is directly applied for the separation of lanthanides from the uranium leach liquor and lime cake. The accuracy of the method is checked by analyzing synthetic mixture containing known amounts of traces of lanthanides and also by comparing with another standard separation procedure like ion exchange method, and the recovery was better than 95%. The method is rapid, simple, accurate and suitable for the separation of lanthanides from uranium, iron and calcium rich materials. The precision of the method is characterized by an RSD of 2 to 4%.     

Bis(2-ethylhexyl)sulfoxide as an extractant for americium(III) from aqueous nitrate media

Solven extraction separation of americium(III) from dilute aqueous nitrate media into n-dodecane by bis(2-ethylhexyl)sulfoxide (BESO) has been investigated over a wide range of experimentgal conditioins. Very poor extractablity of Am(III), necessitated the use of calcium nitrate as the salting-out agent. Effects of certain variables such as acidity, extractant concentration, salting-out agent concentration, organic diluents on the metal extraction by BESO have been examined in detail. By increasing the concentration of BESO in organic phase or calcium nitrate in aqueous phase, nearly quantitative extraction of americium even from moderate acidity is accomplished. Slope analyses applied to Am(III) distribution experiments from acidic nitrate solutions indicate predominant formation of the risolvated organic phase complex, Am(NO₃)₃)·3BESO for which equilibrium constant is found to be, log K_x=1.99. Extraction behavior of Am(III) has also been evlauated in the presence of several water-miscible polar organic solvents to stuy their possible synergistic effects on its extraction. Extractability of americium increased 5 to 10-fold withi increasing conentration of some of these additives, with maximum enhancement being observed in the presence of acetone or acetonitrile. Recovery of BESO from loaded americium is easily obtained using dilute nitric acid as the strippant.     

Recovery of plutonium and americium from laboratory acidic waste solutions using tri-n-octylamine and octylphenyl-N-N- diisobutylcarbamoylmethylphosphine oxide

Plutonium from acidic waste solutions has been recovered quantitatively using tri-n-octylamine (TnOA) in xylene and americium using a mixture of octylphenyl-N-N- diisobutylcarbamoylmethylphosphine oxide (CMPO) and TBP in dodecane by extraction and extraction chromatographic methods. The Pu ( IV ) TnOA species extracted into the organic phase from higher nitric acid concentrations has been confirmed as (R(3)NH)(2)Pu(NO(3))(6) (where R(3)N = TnOA by employing slope analysis as well as spectrophotometric studies.     

Recovery of uranium and thorium from zirconium oxychloride by solvent extraction

A solvent extraction process is proposed to recover uranium and thorium from the crystal waste solutions of zirconium oxychloride. The extraction of iron from hydrochloride medium with P350, the extraction of uranium from hydrochloride with N235, and the extraction of thorium from the mixture solutions of nitric acid and the hydrochloric acid with P350 was investigated. The optimum extraction conditions were evaluated with synthetic solutions by studying the parameters of extractant concentration and acidity. The optimum separation conditions for Fe (III) are recognized as 30% P350 and 4.5 to 6.0 M HCl. The optimum extraction conditions for U (VI) are recognized as 25% N235 and 4.5 to 6.0 M HCl. And the optimum extraction conditions for Th (VI) are recognized as 30% P350 and 2.5 to 3.5 M HNO₃ in the mixture solutions. The recovery of uranium and thorium from the crystal waste solutions of zirconium oxychloride was investigated also. The results indicate that the recoveries of uranium and thorium are 92 and 86%, respectively.     

Separation of neptunium, plutonium, americium and curium from uranium with di-(2-ethylhexyl)-phosphoric acid (HDEHP) for radiometric and ICP-MS analysis

The possibility of using di-(2-ethylhexyl)-phosphoric acid (HDEHP) in solvent extraction for the separation of neptunium, plutonium, americium and curium from large amounts of uranium was studied. Neptunium, plutonium, americium and curium (as well as uranium) were extracted from HNO₃, whereafter americium and curium were back-extracted with 5M HNO₃. Thereafter was neptunium back-extracted in 1M HNO₃ containing hydroxylamine hydronitrate. Finally, plutonium was back-extracted in 3M HCl containing Ti(III). The method separates²³⁸Pu from²⁴¹Am for α-spectroscopy. For ICP-MS analysis, the interferences from²³⁸U are eliminated: tailing from²³⁸U, for analysis of²³⁷Np, and the interference of²³⁸UH⁺ for analysis of²³⁹Pu. The method has been used for the analysis of actinides in samples from a spent nuclear fuel leaching and radionuclide transport experiment.     

Modeling and Flowsheet Design of an Am Separation Process Using TODGA and H4TPAEN

Recycling americium from spent fuels is an important consideration for the future nuclear fuel cycle, as americium is the main contributor to the long-term radiotoxicity and heat power of the final waste, after separation of uranium and plutonium using the PUREX process. The separation of americium alone from a PUREX raffinate can be achieved by co-extracting lanthanide (Ln(III)) and actinide (An(III)) cations into an organic phase containing the diglycolamide extractant TODGA, and then stripping Am(III) with selectivity towards Cm(III) and lanthanides. The water soluble ligand H₄TPAEN was tested to selectively strip Am from a loaded organic phase.Based on experimental data obtained by Jülich, NNL and CEA laboratories since 2013, a phenomenological model has been developed to simulate the behavior of americium, curium and lanthanides during their extraction by TODGA and their complexation by H₄TPAEN (complex stoichiometry, extraction and complexation constants, kinetics). The model was gradually implemented in the PAREX code and helped to narrow down the best operating conditions. Thus, the following modifications of initial operating conditions were proposed:

• •An increase in the concentration of TPAEN as much as the solubility limit allows.
• •An improvement of the lanthanide scrubbing from the americium flow by adding nitrates to the aqueous phase.

A qualification of the model was begun by comparing on the one hand constants determined with the model to those measured experimentally, and on the other hand, simulation results and experimental data on new independent batch experiments.A first sensitivity analysis identified which parameter has the most dominant effect on the process. A flowsheet was proposed for a spiked test in centrifugal contactors performed with a simulated PUREX raffinate with trace amounts of Am and Cm. If the feasibility of the process is confirmed, the results of this test will be used to consolidate the model and to design a flowsheet for a test on a genuine PUREX raffinate. This work is the result of collaborations in the framework of the SACSESS European Project.     

Study on the synergic extraction of uranium(VI) with petroleum sulfoxides (PSO) and tri-n-butyl phosphate (TBP) mixture

The synergic extraction of uranium(VI) from nitric acid solution with petroleum sulfoxides (PSO) and tri-n-butyl phosphate (TBP) mixture has been studied. It has been found that maximum synergic extraction effect occurs if the molar ratio of PSO to TBP is two to three. The composition of the complex of synergic extraction is UO₂(NO₃)₂·TBP·PSO. The formation constant of the complex isK _PT=8.19. The effect of extractant concentration, nitric acid concentration, salting-out agent concentration and temperature on the extraction equilibrium of uranium(VI) was also studied.     

Separation of AM(III) by extraction from nitrate solutions using some quaternary benzylammonium salts

Benzyldimethyldodecylammonium nitrate and benzyltrioctylammonium nitrate were used for the extraction of Am(III) from aqueous nitrate solutions. The dependence of the extraction performance for Am(III) on the concentration of nitric acid, the kind and concentration of salting-out agents in the aqueous phase, and the kind of solvent was investigated. Americium is extracted by the above quarternary salts as a R₄NAm(NO₃)₄ associate. The extraction of Am(III) is compared with the extraction of lanthanides. The high differences in the distribution coefficients for lanthanides and americium can be utilized for the separation of lanthanides and americium.     

Extraction studies of trivalent ions from TALSPEAK medium using phsophonic acid-TBP solvent mixture

Indigenously synthesized extractant, phenyl (octyl) phosphonic acid (POPA) in tri-n-butylphosphate (TBP) and dodecane, has been investigated for the separation of americium from trivalent lanthanides in nitric acid medium as well as diethylene triaminepentaacetic acid (DTPA) and lactic acid mixture (TALSPEAK medium). Various experimental parameters like concentration of DTPA, lactic acid, TBP, nitrate ions and pH of the aqueous feed solution have been optimized to obtain the highest separation factor between americium and europium. Bulk actinide–lanthanide separation reagent, tetra (ethylhexyl) diglycolamide (TEHDGA), was equilibrated with simulated solution of americium and lanthanides, equivalent in concentration to the reprocessing waste originating from PHWR spent fuel. DTPA/lactic acid mixture was used to strip the metal ions from the loaded organic phase and re-extracted into POPA in TBP/dodecane to evaluate the separation factor of individual lanthanides with respect to americium. Very good separation factors between americium and trivalent lanthanides were obtained.     

Extraction of Am(III) from different acid media by di-2-ethyl hexyl phosphoric acid containing phosphorous pentoxide

The extraction of Am(III) from nitric, hydrochloric, oxalic, phosphoric and hydrofluoric acids was studied using 0.4F di-2-ethyl hexyl phosphoric acid (HDEHP) containing 0.1M phosphorous pentoxide (P₂O₅) in dodecane/xylene. The extraction with pure 0.4F HDEHP was found to be negligible from all the media studied. However, the presence of a small amount of P₂O₅ in it increased the extraction substantially. The distribution ratios of Am(III) obtained for HDEHP - P₂O₅ mixture 3M nitric acid containing different concentrations of oxalic acid/phosphoric acid/hydrofluoric acid are in the order of 200-250. The same for 3M hydrochloric acid is very high (800). These distribution ratios are sufficiently high for the quantitative extraction of Am(III) from all the acid media studied. Different reagents such as ammonium oxalate, sodium oxalate, oxalic acid, hydrofluoric acid, sodium carbonate and potassium sulphate were explored for the back extraction of Am(III) from 0.4F HDEHP + 0.1M P₂O₅ in dodecane/xylene. Of these, 0.35M ammonium oxalate and 1M sodium carbonate were found to be most suitable. The back extraction of Am(III) was also attempted with water and 1M H₂SO₄, HNO₃, HClO₄ and HCl solutions after allowing the extracted organics to degrade on its own. It was found that more than 90% of Am could be back extracted with these acids. Using this method more than 90% of Am(III) was recovered from nitric acid solutions containing calcium and fluoride ions.     

Extraction of uranium and transuranium elements with <Emphasis Type="Italic">tert</Emphasis>-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.     

Separation and determination of <Superscript>241</Superscript>Am in urine samples from radiation workers using PC88-A and alpha spectrometry

Bioassay technique is used for the estimation of actinides present in the body based on their excretion rate through body fluids. For occupational radiation workers urine assay is the preferred method for monitoring of chronic internal exposure. Determination of low concentrations of actinides such as plutonium, americium and uranium at low level of mBq in urine by alpha spectrometry requires pre-concentration of large volumes of urine. This article deals with standardization of analytical method for the determination of ²⁴¹Am isotope in urine samples using Extraction Chromatography (EC) and ²⁴³Am tracer for radiochemical recovery. The method involves oxidation of urine followed by co-precipitation of americium along with calcium phosphate. This precipitate after treatment is further subjected to calcium oxalate co-precipitation. Separation of Am was carried out by EC column prepared by PC88-A (2-ethyl hexyl phosphonic acid 2-ethyl hexyl monoester) adsorbed on microporous resin XAD-7 (PC88A-XAD7). Am-fraction was electro-deposited and activity estimated using tracer recovery by alpha spectrometer. Ten routine urine samples of radiation workers were analyzed and consistent radiochemical recovery was obtained in the range 44–60% with a mean and standard deviation of 51 and 4.7% respectively.     

Milking process for the preparation of 239Np

A milking process is described for preparing ²³⁹Np from ²⁴³Am. The process includes the stabilization of Np(IV) with ascorbic acid, isolation of Np(IV) by extraction with TOPO/dodecane and stripping of Np with (NH₄)₂CO₃. The yield amounts to 60%. During milking, about 2% of the ²⁴³Am ends up in the scrub which is collected and reprocessed together with the remaining feed after 25 operations using extraction chromatography.