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%.     

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%.     

Analytical application of DHOA for the determination of trace metallic constituents in U based fuel materials by ICP-AES

DHOA (Di-n-hexyl-octanamide) is one of the alternative extractants to TBP (tri-n-butyl phosphate) known for the extraction of uranium from moderate nitric acid medium without significant extraction of the fission products. Analytical application of DHOA was explored to develop a methodology for determination of trace metallic constituents in uranium based nuclear materials. This involved the separation of uranium matrix by 1.1 M DHOA-dodecane followed by the analysis of the raffinate for trace constituents by Inductively Coupled Plasma Atomic Emission Spectrometry (ICP-AES). A systematic study showed that four contacts of 1.1 M DHOA-dodecane were required for quantitative extraction of U from 4 M HNO₃ feed for the sample size of 1 g in 10 mL. The feasibility of using DHOA for extraction of U from trace metallic constituents in U based fuel materials without losing trace quantities of analytes of interest was studied by using synthetic samples after appropriate spiking of common impurities and critical elements at their required specification limits (common elements—5 ppm, critical elements—1 ppm). A systematic study was carried out to compare the analytical performance of DHOA with TBP, which revealed that DHOA could successfully be employed for the determination of 19 trace constituents with lower estimation limits of 5 ppm for common impurities and 1 ppm for critical elements.     

Synergic extraction for recovery of americium from waste solution containing an excess of uranium

A method based on synergic extraction has been evolved for the recovery of tens of milligrams of americium from analytical wastes in 7-8M HNO₃ medium containing excess uranium as a two step procedure viz., (1) separation of uranium by contacting with TBP in dodecane and (2) recovery of americium by an extraction-cum-strip cycle using a synergic mixture of PMBP-TBP in dodecane after decreasing the acidity of the solution. Other transition metals such as iron found in significant proportion were separated from Am by using the difference in the kinetics of extraction of iron and americium into HPMBP-TBP-dodecane mixture by short duration contacts. About 99% of Am could be recovered into about 20% of its initial volume. This revised version was published online in July 2006 with corrections to the Cover Date.     

Extraction chromatographic studies of metal ions using N,N,N',N'-tetraoctyl diglycolamide as the stationary phase

N,N,N′,N′-tetraoctyl diglycolamide (TODGA) has been used as the stationary phase in the extraction chromatographic separation of actinides and other metal ions from pure nitric acid as well as from simulated high-level waste (SHLW). Chromosorb-W was found to be a better support material amongst the different solid supports evaluated viz. chromosorb-W, chromosorb-102, XAD-4 and XAD-7. Uptake profiles of various metal ions, such as U(VI), Pu(IV), Am(III), Eu(III), Fe(III), Sr(II) and Cs(I) were obtained as a function of acidity by batch studies using TODGA/chromosorb-W. Effect of macro concentration of Nd, Fe and U suggested that the uptake of Am(III) is mainly influenced by the presence of trivalent lanthanide ions. Breakthrough capacity of the resin material for Am(III) in presence of macro amount of Eu(III) was determined in the successive cycles of loading and elution. Loading capacity of the column was found to be 20 mg of Eu/g of the resin material. Elution studies of Am(III) suggested that 0.01 M EDTA was effective amongst different eluents used.     

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%.     

Separation of curium from americium using composite sorbents and complexing agent solutions: part 2

TODGA–PAN composite sorbent and (PhSO₃H)₂–BTPhen in nitric acid solution were employed as a system for separation of curium from americium. The influence of aqueous phase composition (complexing agent and nitric acid concentrations) on weight distribution coefficients and Cm/Am separation factor was studied in batch experiments with trace amounts of ²⁴¹Am and ²⁴⁴Cm. Based on the results obtained, column experiment was designed and conducted. The Cm/Am separation factor of 3.8 ± 0.1 found in batch experiments with TODGA–PAN could be reproduced also in column experiment resulting in good separation of Cm from Am. The efficiency of Cm separation from Am in the TODGA–PAN system was compared with the analogous system with DGA resin (Triskem International). After separation on a 0.5 mL column (φ4.7 × 29 mm) the Cm fraction containing 93% of Cm(III) contained only 3% of Am(III) in optimum conditions.

     

Analysis of Th, U, Pu, and Am in radioactive metal waste using extraction chromatography

In order to analyze actinide elements in radioactive metal waste, the dissolution and chemical separation conditions were optimized. The surfaces of a type 304 stainless steel plate and of pipe waste sampled from the prototype advanced thermal reactor (Fugen) were dissolved in mixed acid solution (HNO₃:HCl:H₂O = 1:1:4). The resulting solution was evaporated to dryness and dissolved with 2 mol/dm³ of HNO₃ to prepare sample solutions. In order to analyze trivalent actinide elements in the sample solution containing a large amount of Fe(III) (>0.1 g) using TRU resin, the effect of Fe(III) concentration on the recovery of Am(III) and reduction effect of Fe(III) to Fe(II) with ascorbic acid were studied. On the basis of results of this study, chemical separation scheme was constructed and Pu and Am in the sample solutions were separated. Thorium and U in the sample solutions were separated with UTEVA resin. High recoveries for all experimented elements were obtained from the analysis of spiked sample solutions, the effectiveness of the method was confirmed.     

Separation and recovery of Cm from Cm–Pu mixed oxide samples containing Am impurity

Curium was separated and recovered as an oxalate from a Cm–Pu mixed oxide which had been a ²⁴⁴Cm oxide sample prepared more than 40 years ago and the ratio of ²⁴⁴Cm to ²⁴⁰Pu was estimated to 0.2:0.8. Radiochemical analyses of the solution prepared by dissolving the Cm–Pu mixed oxide in nitric acid revealed that the oxide contained about 1 at% of ²⁴³Am impurity. To obtain high purity curium solution, plutonium and americium were removed from the solution by an anion exchange method and by chromatographic separation using tertiary pyridine resin embedded in silica beads with nitric acid/methanol mixed solution, respectively. Curium oxalate, a precursor compound of curium oxide, was prepared from the purified curium solution. 11.9 mg of Cm oxalate having some amounts of impurities, which are ²⁴³Am (5.4 at%) and ²⁴⁰Pu (0.3 at%) was obtained without Am removal procedure. Meanwhile, 12.0 mg of Cm oxalate (99.8 at% over actinides) was obtained with the procedure including Am removals. Both of the obtained Cm oxalate sample were supplied for the syntheses and measurements of the thermochemical properties of curium compounds.     

Distribution behavior of U(VI), Am(III) and Eu(III) on diglycolamide based extraction chromatographic resin in perchloric acid medium

An attempt has been made in the present work to investigate the role of anion for the uptake of Am(III)/Eu(III)/U(VI) by extraction chromatography (EXC) resin incorporating tetra-n-octyl-3-oxapentanediamide, commonly referred to as tetra-octyl diglycolamide (TODGA). In contrast to the nitric acid, perchloric acid medium favors extraction of trivalent metal ions even at low acidity (pH 2) and is almost insensitive to the acidity up to 5 M. Exceptionally large distribution coefficients (10⁵–10⁶) in the wide range of perchlorate concentration (10^?2–5 M) is quite unusual and is by far the largest reported in the literature for Am(III)/Eu(III). Thermodynamic data suggests the possibility of inner sphere/cation exchange mechanism involving TODGA aggregates at higher acidity but outer sphere/cation exchange mechanism at low acidity for Eu(III). There is a possibility of employing TODGA based EXC resin for the remediation of liquid waste (contaminated with long lived transuranics like ^241/243Am and ²⁴⁵Cm) in the wide range of acidity.     

A novel partitioning process for treatment of high level liquid waste using macroporous silica-based adsorbents

To separate the long-life and significant fission product elements from high level liquid waste (HLLW), a novel partitioning process for the treatment of HLLW has been studied experimentally based on column separation technique using macroporous silica-based adsorbents. This process consists of (1) Cs and Rb are removed by the first separation column packed with (calix[4] + dodecanol)/SiO₂–P adsorbent; (2) Sr and Ba are eluted out by the second separation column packed with (DtBuCH18C6 + dodecanol)/SiO₂–P adsorbent; (3) Pd is partitioned by the third separation column packed with MOTDGA–TOA/SiO₂–P adsorbent; (4) Ru, Rh and Mo can be separated by the fourth separation column packed with TODGA/SiO₂–P adsorbent; (5) Am is separated from RE by the fifth column is packed with isobutyl-BTP/SiO₂–P adsorbent. The experimental results indicated that this partitioning process is essentially feasible.     

Radiation stability of several polymeric supports used for radionuclide transport from nuclear wastes using liquid membranes

In an attempt to evaluate radiation stability of several polymeric materials used as the support in supported liquid membrane studies for the transport of radionuclides from nuclear waste, flat sheets made from polytetrafluoroethylene, polysulfone, polyether sulfone, polyacrylonitrile and polyvinylidenefluoride were irradiated to varying extents using a ⁶⁰Co gamma ray source and subsequently, the transport efficiency of the irradiated flat sheets were evaluated. The membrane integrity was assessed from the transport rates of Am³⁺ from a feed containing 3 M HNO₃ into a receiver phase containing 0.01 M HNO₃ as the strippant while 0.1 M TODGA (N,N,N′,N′-tetraoctyldiglycolamide) + 0.5 M DHOA (di-n-hexyloctanamide) in n-dodecane was used as the carrier extractant. The radiation stability of the membrane filters was evaluated after irradiating them up to 20 MRad absorbed dose in a gamma chamber.     

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.     

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.     

Comparative evaluation of two substituted diglycolamide extractants for ‘actinide partitioning’

Extraction behaviour of actinides, lanthanides, fission products and structural elements has been studied with the two diglycolamide extractants, namely N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (T2EHDGA) and N,N,N′,N′-tetraoctyl diglycolamide (TODGA). The acid extraction studies suggested that T2EHDGA (K_H: 1.8) is less basic as compared to its linear homologue, TODGA (K_H: 4.1). The distribution ratio of Am(III) by 0.1 M diglycolamides followed the order: TODGA > T2EHDGA. The number of ligand molecules present in the stoichiometry of the extracted species of Am(III) was found to be three and four for T2EHDGA and TODGA, respectively. Thermodynamics studies suggested that the extraction of Am(III) by both the extractants is exothermic in nature. The radiolytic stability of TODGA and T2EHDGA solutions in n-dodecane has been investigated. Due to lower distribution ratio of Am by T2EHDGA, 0.2 M of its solution has been used as compared to 0.1 M solution of TODGA. The distribution behaviour of various metal ions, viz. Am, Nd, Fe, Mo, Cr, Sr and Cs has been studied from nitric acid as well as from simulated high level waste solution.     

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.     

Adsorption and desorption behavior of tetravalent zirconium onto a silica-based macroporous TODGA adsorbent in HNO3 solution

To understand the separation behavior of Zr(IV) in the partitioning process for high level liquid waste, a silica-based macroporous adsorbent (TODGA/SiO₂-P) was prepared by impregnating N,N,N′,N′-tetraoctyl-3-oxapentane-1,5-diamide (TODGA) into a macroporous silica/polymer composite particles support (SiO₂-P). Adsorption and desorption behavior of Zr(IV) from nitric acid solution onto silica-based TODGA/SiO₂-P adsorbent were investigated by batch experiment. It was found that TODGA/SiO₂-P showed strong adsorption affinity to Zr(IV) and this adsorption process reached equilibrium state around 6 h at 298 K. Meanwhile, HNO₃ concentration had no significant effect on the adsorption of Zr(IV) above 1 M. From calculated thermodynamic parameters, this adsorption process could occur spontaneously at the given temperature and was confirmed to be an exothermic reaction. This adsorption process could be expressed by Langmuir monomolecular layer adsorption mode and the maximum adsorption capacity were determined to be 0.283 and 0.512 mmol/g for Zr(IV) at 298 and 323 K, respectively. In addition, more than 90 % of Zr(IV) adsorbed onto adsorbent could be desorbed with 0.01 M diethylenetriamine pentaacetic acid solution within 24 h at 298 K.     

Separation of Am(III) from nitric acid medium by phosphonium ionic liquid–hydroxyacetamide mixture

A strongly hydrophobic phosphonium ionic liquid, trihexyltet radecylphosphonium bis(trifluoromethanesulfonyl)imide ([P₆₆₆₁₄][NTf₂]) was employed as the diluent for the extraction behavior of Am(III) using N,N-dihexyl-2-hydroxyacetamide(DHHy) as extractant. The extractibility of americium(III) in [P₆₆₆₁₄][NTf₂] phase was measured as a function of various parameters such as aqueous phase acidity (0.1–8 M), extractant concentration (0.01–0.15 M), equilibration time (5–120 min) and temperature (298–333 K). The extraction performance observed in DHHy/[P₆₆₆₁₄][NTf₂] was compared with those observed in N,N-dihexyloctamide (DHOA) in [P₆₆₆₁₄][NTf₂] and DHHy in other diluents such as [C₄mim][NTf₂] and n-dodecane. The effect of temperature on D _Am(III) in ionic liquid system and recovery of Am(III) from the loaded phase were ascertained in detail.     

Assay of actinides in human urine by rapid method

A method using DGA resin (N,N,N′,N′-tetra-n-octyldiglycolamide on an inert support) was developed for the rapid analysis of actinides in urine samples. Samples acidified with HCl to 4 M were loaded directly (without digestion) onto a DGA column. Actinides were stripped simultaneously, α-sources were prepared by co-precipitation with NdF₃. Americium, plutonium and uranium were separated with acceptable high recoveries (40–80%). The americium, plutonium and uranium content of 100–200 ml urine samples was determined within 24 h with detection limits as low as 0.01 Bq l^?1. Based on model experiments using ¹⁴C-spiked urea, it was proven that high urea content can affect americium separation deleteriously due to irreversible fixing of americium on DGA resin.     

The separation of241Am from bones by coprecipitation with bismuth phosphate

The conditions of²⁴¹Am separation from bone by coprecipitation with BiPO₄ were studied. It was found that by coprecipitation with BiPO₄ ²⁴¹Am can be separated with high yield from different amount of bone. The main condition of the achievement of a high yield is a low Fe/III/ concentration in solution at americium coprecipitation.