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 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.     

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 and extraction chromatographic separation of minor actinides from sulphate bearing high level waste solutions using CMPO

Bench-Scale studies on the partitioning and recovery of minoractinides from the actual and synthetic sulphate-bearing high level waste (SBHLW) solutions have been carried out by giving two contacts with 30% TBP to deplete uranium content followed by four contacts with 0.2M CMPO+1.2M TBP in dodecane. The acidity of the SBHLW solutions was about 0.3M. In the case of actual SBHLW, the final raffinate contained about 0.4% -activity originally present in the HLW, whereas with synthetic SBHLW the -activity was reduced to the background level.¹⁴⁴Ce is extracted almost quantitative in the CMPO phase,¹⁰⁶Ru about 12% and¹³⁷Cs is practically not extracted at all. The extraction chromatographic column studies with synthetic SBHLW (aftertwo TBP contacts) has shown that large volume of waste solutions could be passed through the column without break-through of actinide metal ions. Using 0.04M HNO₃>99% Am(III) and rare earths could be eluted/stripped. Similarly >99% Pu(IV) and U(VI) could be eluted.stripped using 0.01M oxalic acid and 0.25M sodium carbonate, respectively. In the presence of 0.16M SO ₄ ^2– (in the SBHLW) the complex ions AmSO ₄ ⁺ , UO₂SO₄, PuSO ₄ ²⁺ and Pu(SO₄)₂ were formed in the aqueous phase but the species extracted into the organic phase (CMPO+TBP) were only the nitrato complexes Am(NO₃)₃·3CMPO, UO₂(NO₃)₂·2CMPO and Pu(NO₃)₄·2CMPO. A scheme for the recovery of minor actinides from SBHLW solution with two contacts of 30% TBP followed by either solvent extraction or extraction chromatographic techniques has been proposed.     

Use of complexones for the extraction separation of rare earth and transplutonium elements with amines

The extraction distribution and separation of rare earth elements and americium from the concentrated lithium nitrate solution with solutions of tertiary amines in organic solvents has been studied as a function of the composition and structure of complexones of the polyaminepolyacetic acid series by a radioactive tracer method. It has been found that diethylenetriaminepentaacetic acid is suitable for the separation of REE from americium(III). The apparent stability constants for the lanthanide complexes with EDTA and DTPA in concentrated litium nitrate solutions have been obtained by extraction, pH-metric titration and solubility. Using these constants, the optimum conditions of separation have been found and the separation factors of REE calculated. The calculated and experimental values are in good agreement. The optimum conditions for the separation of americium(III) from REE in a wide range of lanthanide and complexone concentrations (10⁻¹–10⁻⁶ M) have been determined.     

Studies on the partitioning of actinides from high level liquid waste solution employing supported liquid membrane with Cyanex-923 as carrier

The present paper deals with the studies on the partitioning of actinides from high level liquid waste solution of PUREX origin employing supported liquid membrane technique. The process uses solution of Cyanex-923 in n-dodecane as a carrier with poly tetra fluoro ethylene support and a mixture of citric acid, formic acid and hydrazine hydrate as a receiving phase. Transport studies are carried out for ²⁴¹Am under different experimental conditions to optimize the transport parameters such as feed acidity, carrier concentration, effect of uranium, trivalent metal ion and salt concentration in the feed. Studies indicated good transport of actinides across the membrane from nitric acid medium. Under the optimized conditions the transport of ²⁴¹Am is studied from a uranium depleted synthetic PHWR-HLLW and finally the technique has been used for the partitioning of alpha emitters from an actual research reactor-HLLW. High concentration of uranium in the feed is found to retard the transport of americium, suggesting the need for prior removal of uranium from the waste.     

Recovery of neptunium from highly radioactive waste solutions of PUREX origin using CMPO

The partitioning and recovery of²³⁷Np from three types of simulated high level waste solutions originating from PUREX processing of spent nuclear fuels such as sulfate bearing high level waste (SB-HLW), HLW from a pressurised heavy water reactor (PHWR-HLW) and from a fast breeder reactor (FBR-HLW) have been carried out using a mixture of 0.2M CMPO and 1.2M TBP in dodecane. Quantitative extraction of neptunium was possible by either oxidizing it to the hexavalent state keeping K₂Cr₂O₇ at 0.01M concentration or by reducing it to tetravalent state keeping Fe²⁺ at 0.02M concentration. Stripping of neptunium was carried out using different reagents, such as dilute nitric acid, oxalic acid and sodium carbonate. Almost quantitative recovery of neptunium has been achieved during these studies.     

Separation of carrier-free 90Y from high level waste by supported liquid membrane using KSM-17

A generator system has been developed for the preparation of carrier-free ⁹⁰Y from ⁹⁰Sr present in the high level waste (HLW) of the Purex process by employing a supported liquid membrane (SLM) using 2-ethylhexyl-2-ethylhexyl phosphonic acid (KSM-17 equivalent to PC 88A) supported on a polytetrafluoro ethylene (PTFE) membrane. When uranium depleted Purex HLW at appropriate acidity is passed sequentially through octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) sorbed on chromosorb-102 (abbreviated as CAC) and Zeolite AR1 (synthetic mordenite) columns, all the trivalent, tetravalent and hexavalent metal ions and monovalent ¹³⁷Cs ions are sorbed. After adjusting to pH 2 with NaOH the resulting effluent is used as feed in a single stage membrane cell partitioned with a PTFE membrane impregnated with KSM-17 and having a feed and receiver compartment with 5.0 ml capacity each. The receiver compartment was filled with a 0.5M HNO₃ or 0.5M HCl stripping solution. ⁹⁰Y alone is preferentially transported across the membrane leaving behind all the impurities viz. ⁹⁰Sr, ¹²⁵Sb, ¹⁰⁶Ru, ¹⁰⁶Rh, etc. in the feed compartment. This technique can yield ⁹⁰Y in mCi levels in a pure and carrier-free form for medical applications. The feed can be reused repeatedly after allowing for ⁹⁰Y buildup.     

Recovery of actinides extracted by Truex solvent from high level waste using complexing agents

This work deals with the batch studies on stripping of actinides extracted by a mixture octyl(phenyl)-N,N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and tri-n-butyl phosphate (TBP) in n-dodecane (Truex solvent) from simulated high level waste (HLW) solution. The stripping of americium and plutonium from acid-bearing CMPO-TBP mixture is carried out using a mixture of weak acid, weak base and complexing agent as strippant. A mixture of formic acid, hydrazine hydrate and citric acid appeared to be best suited for efficient stripping of americium and plutonium. With appropriate modifications in the concentration of individual constitutents, this strippant can be used for the recovery of actinides from loaded Truex solvent with any acid content.     

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.     

Preparation, characterization and application of thorium oxalate for sorption of plutonium and americium from aqueous solution

Synthetic inorganic exchangers exhibit good thermal and radiation stability. Thorium oxalate precipitate shows potential for co-precipitation of plutonium and americium from oxalate supernatant generated during plutonium oxalate precipitation. In the present study, efforts were made to prepare thorium oxalate precipitate to be used for column operation. Distribution ratios were determined to optimize conditions for sorption of plutonium and americium on thorium oxalate from nitric acid + oxalic acid solutions with composition similar to that of oxalate supernatant. Column experiments were also performed to evaluate the sorption capacity of thorium oxalate for plutonium and americium from the same medium. The result showed that, thorium oxalate prepared in 1.75M HNO₃ at 70 °C is suitable for column operations. These studies showed that plutonium and americium could be simultaneously removed from aqueous solutions with composition similar to plutonium oxalate waste using glass column packed with thorium oxalate and these nuclides could be recovered by eluting with 3M HNO₃.     

Transport of99mTc across TBP-kerosene oil supported liquid membranes

Transport of^99mTc across tri-n-butylphosphate (TBP) kerosene oil supported liquid membranes (SLM) has been studied under various conditions. Presence of dichromate ions helps avoid activity scavenging effects. Concentration increase of TBP, the complexing carrier used in the present study has a positive effect on flux (J) and permeability (P) of these ions, as up to 2.87M there is an increase in J and P values. HCl concentration in the feed solution increases J and P with their maximum values at 2.5–3.0M HCl in the feed. Above this concentration there is a decrease in flux and permeability of^99mTc(VII) ions. The given ions are stripped with LiCl or NaCl solutions but more with NaOH. The optimum conditions of transport of the given ions are 2.5M HCl concentration in the feed, 2.87M TBP concentration in the membrane and 1M NaOH concentration in the strip solution. Equations have been developed to indicate the relation between flux, J, viscosity, of TBP in organic membrane phase, temperature, T, [H⁺], in the aqueous feed solutions and Tc ion concentration in the feed solution. Based on P, the values determined from liquid membrane experiments, the quantitative flux values of Tc(VII) ions were also determined as a function of TBP concentration in the membranes, and HCl and Tc concentration in the feed solution using the given equations. This experimental technique provides quantitative results from trace level activity transfer experiments.     

Enhanced selectivity for actinides over lanthanides with CMPO ligands secured to a C(3)-symmetric triphenoxymethane platform

A ligand system containing three preorganized carbamoylmethylphosphine oxide (CMPO) moieties anchored onto a rigid C(3)-symmetric triphenoxymethane platform has been developed for facile metal complexation and subsequent extraction from aqueous acidic nuclear waste streams. Intended to mimic the 3:1 CMPO-actinide stoichiometry of the extracted species in the TRUEX nuclear waste treatment process, the CMPO arms on this ligand are oriented such that all three CMPO moieties can cooperatively bind a metal ion. Extractions of simulated nuclear waste streams (10(-4) M metal in 1 M nitric acid) with solutions of this ligand in methylene chloride (10(-3) M) reveal a high affinity for the actinide thorium and a very low, but constant, affinity for the lanthanides across the series. Thorium and five lanthanide (lanthanum, cerium, neodymium, europium, and ytterbium) nitrate complexes of this ligand have been synthesized and fully characterized by X-ray crystallography, (1)H and (31)P NMR spectra, and FT-ICR-MS to elucidate the mechanism of this unique actinide selectivity. All six oxygen donors from the three CMPO arms of the ligand and one or two nitrate counterions coordinate these metals to afford 2+ cationic complexes in every case. Because of the large size of the ligand, both the thorium and lanthanide complexes present similarly charged and sized surfaces to the extraction solvents, but the thorium complex is extracted quantitatively over the lanthanide complexes. A possible rationale for this extraction behavior difference is presented and further illustrated by the extraction properties of this ligand system for the alkali metals (lithium, sodium, potassium, rubidium, and cesium) as picrate salts and by the solid- and solution-state structures of its lithium picrate complex.     

Synthesis and application of a macroporous silica-based polymeric octyl(phenyl)-N,N-diisobutylcarbamoylmethylphoshine oxide composite for the chromatographic separation of high level liquid waste

Summary The Minor Actinides Recovery from HLW by Extraction Chromatography (MAREC) process was used mainly for the separation of minor actinides (MAs) and some specific fission products (FPs) from highly active liquid waste (HLW) by the composite CMPO/SiO₂-P of the macroporous silica based polymeric octyl(phenyl)-N,N-diisobutylcarbamoylmethylphoshine oxide (CMPO) and others. In this study a cascade of chromatographic separation was performed on a 3.0M HNO₃ solution containing 5.0 ^. 10^-3M of 13 elements, at 323 K. The cascade consisted of three columns the first and second ones were packed with CMPO/SiO₂-P and the third with SiO₂-P particles. The first column was employed to prepare various eluents containing saturated CMPO. The second column was used for separation into groups. The CMPO of CMPO/SiO₂-P was recovered from the effluent by the third column and a CMPO-free effluent containing minor actinides was obtained. The elements contained in the simulated HLW of 3.0M HNO₃ were separated into (1) a non-adsorption group (Sr, Cs, and Ru etc.), (2) a MA-hRE (heavy rare earth)-Mo-Zr group, and (3) a lRE (light rare earth) group by eluting with 3.0M HNO₃, 0.05M DTPA (diethylenetriaminepentaacetic acid) (pH 2.0) and HNO₃ (pH 3.5), respectively. The resultant MA-hRE-Mo-Zr mixture containing minor actinides was then separated into the groups (1) Pd-Ru, (2) MA-hRE, and (3) Mo-Zr by utilizing 3.0M HNO₃, distilled water, and 0.05M DTPA (pH 2.0) as eluents. More than 92% of CMPO in the MA-hRE containing effluent was adsorbed by SiO₂-P particles. The effectivity and technical feasibility of MAREC process were demonstrated.     

Coupled transport of Ag(I) ions through triethanolamine–cyclohexanone-based supported liquid membranes

Facilitated transport of silver(I) ions in acidic medium, across a supported liquid membrane (SLM) by using triethanolamine (TEA) as carrier, dissolved in cyclohexanone, has been investigated. The parameters studied are HNO₃ concentration variation in the feed, pH of the feed solution, carrier concentration in the membrane phase, silver(I) ions concentration in the feed phase and KCN concentration in the stripping phase. Increase in H⁺ concentration by increasing HNO₃ concentration from 0.5 to 1 M results into an increase in silver ions flux but a decrease in flux has been found beyond 1 M HNO₃ concentration in the feed, providing a maximum flux of 3.21 × 10⁻⁷ mol/m² s at 1 M HNO₃. Increase in TEA concentration inside the membrane enhances flux with its maximum value at 2.25 M TEA. Further increase in the concentration of TEA leads to a decreased rate of transport due to the increase in viscosity of membrane liquid. The optimum conditions for Ag(I) ions transport are 1 M HNO₃ (feed), 2.25 M TEA (membrane) and 1.5 M KCN in the stripping phase. It has been observed that Ag(I) flux across the membrane tends to increase with increase in Ag(I) ions concentration in the feed phase. Applying the studied conditions to silver plating waste solutions, Ag ions have been removed up to 99% in a time interval of 5 h.     

Modified <Emphasis Type="Italic">Rhizopus arrhizus</Emphasis> biomass for sorption of <Superscript>241</Superscript>Am and other radionuclides

The improvement and the refinement of non-viable Rhizopus arrhizus biomass were investigated via immobilization. Immobilization was carried out by using sodium alginate/CaCl₂ solution and formaldehyde/HCl cross-linking with dead Rhizopus arrhizus biomass and were used for the sorption of radionuclides from low level effluent wastes. The sodium alginate/CaCl₂ immobilized biomass (ratio 1:2) showed about 86% sorption for ²⁴¹Am activity but due to its soft nature and tendency to undergo distortion in shape, is unsuitable for practical applications. The biomass cross-linked with 15% formaldehyde/0.1 M HCl solution has a relatively high mechanical strength and rigidity. It was showing a sorption of >99% for ²⁴¹Am activity and has the sorption capacity of ~65 mg/g for americium and uranium. Hence, it can be utilized for the removal of radionuclides from radioactive waste effluents.     

Recovery of plutonium from phosphate containing aqueous analytical waste solutions using macroporous anion exchange resin

Distribution ratios of Pu(IV) between 7.5M HNO₃+0.75M H₃PO₄+0.3M H₂SO₄ media and a macroporous anion-exchange resin Amberlyst A-26 (MP) increased from 40 to 250 when 1M aluminium nitrate was added to the aqueous medium. When 1M ferric nitrate was used in place of aluminium nitrate the distribution ratio further increased to 850. The 10% Pu(IV) breakthrough capacities with a 5 ml bed resin column, using synthetic feed solutions containing 1M aluminium nitrate, were 1.4 g l^–1, 3.2 g l^–1 at flow rates of 30 ml per hour and 10 ml per hour, respectively. The corresponding 10% Pu(IV) breakthrough capacities in the presence of 1M ferric nitrate were 8.5 g l^–1 and 12.8 g l^–1. More than 97% of plutonium could be recovered from actual analytical phosphate waste solutions.     

Kinetics of americium(VI) mass transfer through solid supported liquid membrane with HDEHP

The main regularities of membrane extraction of americium under conditions of different redox potentials in aqueous phases have been studied. The physico-chemical model of the process including steps of americium oxidation in feed solution, extraction by membrane, partial reduction on membrane surface, trans-membrane diffusion and reextraction to strip solution has been developed. The calculation of reduction rate constant on membrane surface has been carried out.     

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.     

Liquid-liquid extraction separation and sequential determination of plutonium and americium in environmental samples by alpha-spectrometry

A procedure is described by which plutonium and americium can be determined in environmental samples. The sample is leached with nitric acid and hydrogen peroxide, and the two elements are co-precipitated with ferric hydroxide and calcium oxalate. The calcium oxalate is incinerated at 450 degrees and the ash is dissolved in nitric acid. Plutonium is extracted with tri-n-octylamine solution in xylene from 4M nitric acid and stripped with ammonium iodide/hydrochloric acid. Americium is extracted with thenoyltrifluoroacetone solution in xylene at pH 4 together with rare-earth elements and stripped with 1M nitric acid. Americium and the rare-earth elements thus separated are sorbed on Dowex 1 x 4 resin from 1M nitric acid in 93% methanol, the rare-earth elements are eluted with 0.1M hydrochloric acid/0.5M ammonium thiocyanate/80% methanol and the americium is finally eluted with 1.5M hydrochloric acid in 86% methanol. Plutonium and americium in each fraction are electro-deposited and determined by alpha-spectrometry. Overall average recoveries are 81% for plutonium and 59% for americium.