Purification and recovery of plutonium from uranium oxide obtained in the fast reactor fuel reprocessing plant using hydroxyurea as reductant

Spent fuel discharged from Fast Breeder Test Reactor (FBTR) in Kalpakkam is being reprocessed by modified plutonium uranium reduction extraction (PUREX) process using 30% TBP (tributylphosphate) as extractant in the presence of heavy normal paraffin (HNP) as diluent. Partitioning of uranium (U) and plutonium (Pu) is carried out using oxalate precipitation method. Uranium oxide product obtained by this method contains appreciable amount of plutonium which has to be recovered. Recovery of plutonium from this uranium oxide product is carried out by reducing Pu to inextractable Pu(III) using hydroxyurea (HU) and then uranium is extracted into 30% TBP. A small amount of Pu which is extracted in the organic phase is stripped back to aqueous phase by scrubbing with scrubbing agent containing 0.1 M HU in 4 M nitric acid. Similarly U and Pu are co-extracted into 30% TBP and then Pu is removed by scrubbing with 0.1 M HU in 4 M nitric acid. Further decontamination from Pu is obtained in the stripping stages. By this method Pu contamination in the uranium oxide is brought from 7300 ppm to 0.4–3 ppm (wt/wt). This uranium product obtained can be handled on table top.     

Distribution and identification of Plutonium(IV) species in tri-n-butyl phosphate/HNO3 extraction system containing acetohydroxamic acid

There was a significant research progress achieved with the aim to modify conventional PUREX process by stripping of plutonium from the tri-n-butyl phosphate (TBP) extraction product in the form of non-extractable complexes upon addition of back-hold complexation agents. The present paper reports effects of such salt-free complexant, acetohydroxamic acid (HAHA), on distribution ratio of Pu(IV) under wide concentration of nitric acid and additional nitrate. General formula of plutonium species present in the organic phase can be described as Pu(OH)_x(AHA)_y(NO3)_4−x−y·2TBP·wHNO₃.     

Studies on purification of plutonium from silver ions

Precipitation and solvent extraction methods have been investigated for the purification of plutonium from silver from the solution generated during oxidative dissolution of plutonium oxide using Ag(II) ions. Initial experiments have been carried out using thorium as representative of plutonium. Selecting the optimum conditions, the experiments were repeated with plutonium. The results revealed that Pu can be purified from silver ions either by precipitating silver as silver chloride or silver metal followed by Pu(IV) oxalate precipitation or by selective extraction of Pu(IV) into 20% Aliquat-336 or 30% TBP.     

Hot Test in Mixer Settlers of Alpha Barrier with AHA in PUREX Process

In the PUREX process, the first U-Pu purification cycle (1CUPu) is not efficient enough for the decontamination of uranium flow out of neptunium. In this context, molecules known for their strong complexing power for actinides(IV) in aqueous phase, such as acetohydroxamic acid (AHA) have been tested in batch experiments to strip Np and Pu from TBP solvent loaded with U. A phenomenological model was developed and with the help of this model, a flowsheet of a counter-current alpha barrier process was designed and tested in C17 glove boxes in ATALANTE facility. A decontamination factor DFU/Np of 480 was obtained, higher than DFU/Np required by UNIREP standards.     

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied. The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.     

A novel Np–Pu separation procedure based on extraction with di-(chlorophenyl)-dithiophosphinic acid in nitric solution

The extraction behavior of Pu(III), Pu(IV), Np(IV) and Np(V) with di(chlorophenyl)-dithiophosphinic acid (DCPDTPA) in toluene from nitric acid solutions was studied systematically. In aqueous solution with high nitric acid concentration, the extraction capability (represented by distribution ratio D) for Pu and Np in different valences with DCPDTPA comes as D _Np(IV) > D _Pu(IV) > D _Np(V) > D _Pu(III). A new radiochemical procedure for Np/Pu separation based on DCPDTPA extraction was proposed and tested with simulated samples. The recoveries of Np and Pu are as high as 80 % after the whole separation procedure, with the decontamination factor of trivalent lanthanide fission product element (e.g. Eu) greater than 1.5 × 10⁴. The decontamination factor of Pu–Np is 2.0 × 10³, while the decontamination factor of Np–Pu is greater than 4.8 × 10³ after additional purification.     

Reverse isotope dilution alpha spectrometry using plutionium-239 spike for determining plutonium concentration in high burn-up fuel samples without purification from americium-241 and a bulk of other impurities

A reverse isotope dilution alpha spectrometric /R-IDAS/ method using²³⁹Pu as a spike is described for the determination of plutonium concentration in high burn-up fuel samples wth²³⁸Pu/(²³⁹Pu+²⁴⁰Pu) alpha activity ratio >0.5, without resorting to any purification from²⁴¹Am and a bulk of other impurities. It involves the addition of a pre-clibrated spike solution to a known aliquot of the plutonium sample solution followed by source preparation using TEG as a spreading agent. The results obtained on a number of plutonium samples containing 20–80% of²⁴¹Am /alpha activity wise/ using this method are compared with those achieved by R-IDAS using purification with TTA, with respect to precision and accuracy. Precision and accuracy of 0.5% are demonstrated. This method eliminates the need of any separation and purification of plutonium from²⁴¹Am and a bulk of other impurities like uranium.     

Solid-phase extraction of plutonium in various oxidation states from simulated groundwater using N-benzoylphenylhydroxylamine

Solid-phase extraction of plutonium in different individual and mixed oxidation states from simulated groundwater (pH 8.5) was studied. The extraction of plutonium species was carried out in a dynamic mode using DIAPAK C16 cartridges modified by N-benzoylphenylhydroxylamine (BPHA). It was shown that the extent of recovery depends on the oxidation state of plutonium. The extraction of Pu(IV) was at the level of 98–99% regardless of the volume and flow-rate of the sample solution. Pu(V) was extracted by 90–95% and 75–80% from 10- and 100-mL aliquots of the samples, respectively, whereas the extraction of Pu(VI) did not exceed 45–50%. An equimolar mixture of Pu(IV), Pu(V), and Pu(VI) was extracted by 74%. The distribution coefficients (K _d) and kinetic exchange capacities (S) of plutonium in various oxidation states were measured. It was found that during the sorption process, Pu(V) was reduced to Pu(IV) by 80–90% after an hour-long contact with the solid phase. Pu(VI) is reduced to Pu(V) by 34% and to Pu(IV) by 55%. In the case of mixed-valent solution of plutonium, only Pu(V) and Pu(IV) were found in the effluents.     

Ion exchange behavior of plutonium(IV) on imidazolium nitrate immobilized resin

Imidazolium nitrate anchored on poly(styrene-divinylbenzene) co-polymer, Im-NO₃, has been synthesized and evaluated for plutonium purification. The results are compared with those obtained using Dowex 1 × 4 anion exchange resin. The distribution coefficient (K_d) of Pu(IV) increased with increase in concentration of nitric acid, reached a maximum at 8 M, followed by decrease in K_d values. Rapid ion exchange of Pu(IV) followed by the establishment of equilibrium occurred within 100 min of equilibration and the data was fitted in to first order rate equation. Variation of distribution coefficient of Pu(IV) as a function of exchange capacity and nitrate ion concentration suggest the involvement of anion exchange mechanism is responsible for extraction. The apparent ion exchange capacity was 310 mg/g at 8 M nitric acid. The performance of the Im-NO₃ under dynamic condition was assessed by column breakthrough experiments. Radiolytic degradation of Im-NO₃ resin in presence and absence of nitric acid (8 M) was studied and the results are reported in this paper.     

Separation of137Cs and90Sr from mineralizates of biological materials by dicarbolide of cobalt

The distribution of some radionuclides in the course of¹³⁷Cs and⁹⁰Sr extraction and scrubbing between organic and water phase was determined.¹³⁷Cs and⁹⁰Sr were isolated from the mixture of radionuclides in mineralized biological materials. Dicarbolide of cobalt i. e. 3,3′-commo-bis[undecahydro-1,2-dicarbo-3-closo-dodecaborate] was used as an extracting agent. Quantities of the extracted radionuclides were determined by gamma spectrometric technique. Single and repeated extraction of⁹⁰Sr with 0.01M resp. 0.1M dicarbolide of cobalt in nitrobenzene and scrubbing of coextracted radionuclides by 0.5M HNO₃ were studied. The extraction of⁹⁰Sr was investigated from solutions of a hydrofobizing agent in the same way. Finally, the quantitative extraction of¹³⁷Cs followed by the extraction of⁹⁰Sr from mixtures of radionuclides in a mineralized biological material was studied. Extraction yields from dry and wet mineralizates of biological tissues, from urine and milk were compared. Suitable working conditions for the separation procedures were selected.     

Purification of multiwalled carbon nanotubes by annealing and extraction based on the difference in van der Waals potential

The potential energies of van der Waals interactions between two multiwalled carbon nanotubes (MWNTs) as well as two carbon nanoparticles (CNPs) were calculated and compared on the basis of the continuum Lennard-Jones model. The well depth of the potential is 1 order of magnitude higher for MWNTs than for CNPs, indicating that MWNTs and CNPs can be separated from each other through polymer-induced steric stabilization. On the basis of this prediction, a novel method for the purification of MWNTs was proposed. The method involves a high-temperature annealing (2600 degrees C, 1 h) followed by an extraction treatment with a selected dispersing agent. While the annealing process evaporates the metal particles, the extraction treatment removes CNPs. The quality of the nanotubes obtained after purification was examined by laser Raman, thermogravimetric analysis, and electron microscopy observations.     

Prozessüberwachung eines chromatographischen Verfahrens der Kerntechnik

A chromatographic separation can be controlled only by the determination of the separated component concentration in the column effluent. This control is implemented by an in-line scintillation counting procedure during an extraction chromatographic purification of predecontaminated Th-containing fuel solutions from Np and Pu. In addition, the throughput ofthe process feed solution is determined, and the measuring signals of both determinations are used by appropriated control circuits to regulate the process operation and to perceive the process incidents.      

Evaluation of plutonium(IV) extraction rate between nitric acid and tri‐n‐butylphosphate solution using a glass chip microchannel

Extraction of Pu(IV) with tri‐n‐butylphosphate is performed using a glass chip microchannel to evaluate the extraction rate. Two‐phase flow forms in the microchannel by introducing a solution of Pu(IV) and tri‐n‐butylphosphate with flow rates above 5 μL/min. The Pu(IV) extraction reaction proceeds at the interface between the two phases. To evaluate the extraction rate, the contact time between the two phases is varied from 0.48 to 4.8 s by changing the confluent length of the microchannel and the flow rate. The Pu concentration of each phase collected from the microchannel is measured with an alpha liquid scintillation counter, and the contact time dependence of Pu(IV) extraction is obtained. An extraction model based on diffusion in the microchannel and the reaction at the interface is proposed and applied to determine the extraction rate. The extraction process is assumed to follow pseudo‐first‐order kinetics, and the extraction rate constant of Pu(IV) is determined to be 1.5 × 10^?2 cm/s. The investigation demonstrates that a microfluidic device can be a new tool to determine Pu(IV) extraction rates.     

New data on decomposition of nitric solutions of acetohydroxamic acid

Acetohydroxamic acid (AHA) is an important complexant/reductant for Pu(IV) in the UREX process. It decomposes in the presence of nitric acid. In literature, its decomposition kinetics in nitric acid is traditionally reported as pseudo-first order reaction. In this study, new experimental data were reported for kinetics experiments under wide consecration conditions. It was found that the decomposition reaction was first order with respect to both the components hence overall second order.     

串级萃取量优化理论和精确计算优化萃取比新方程

稀土;串级萃取量优化理论和精确计算优化萃取比新方程     

Modeling of Pu(VI) distribution coefficients in 30 % TBP based PUREX process

Hexavalent plutonium (Pu(VI)) is an important solute in the PUREX (plutonium uranium extraction) process. In 30 % TBP based PUREX solvent extraction system, distribution coefficient of Pu(VI) is much lower than that of Pu(IV). This lower distribution coefficient of Pu(VI) may cause unexpected Pu loss during primary HA extraction in low acid flowsheets. An empirical model for Pu(VI) distribution coefficients in 30 % TBP and its temperature dependency has been reported in this paper. Comparison with literature data revealed a reasonably good agreement between the reported experimental and model predicted values.     

Purification of Bioactive Lipopeptides Produced by Bacillus subtilis Strain BIA

Bacillus subtilis strain BIA was used for the production of bioactive lipopeptides. Different extraction and purification methods were assayed as liquid–liquid extraction, and acid and ammonium sulfate precipitation followed by TLC, SPE, and gel filtration. Active fractions were further purified using RP-HPLC. The molecular mass of the purified product from HPLC was determined through Tris-Tricine SDS-PAGE and MALDI–TOF-MS. The results revealed that Bacillus subtilis strain BIA produced surfactin and iturin like compounds. Coproduction of surfactin and iturin like compounds by this strain is a remarkable trait for a potential biocontrol agent. This paper also includeds techniques that have been developed for the optimal and convenient extraction of bioactive lipopeptides from microbial origin.

     

Radiological characterization of low-and intermediate-level radioactive wastes

An analytical procedure for the determination of activation products ²³⁸Pu, ²⁴¹Pu, ²³⁹Pu/²⁴⁰Pu, ²⁴¹Am, ²³⁷Np, and a fission product ⁹⁰Sr in radioactive wastes is presented. Samples were decomposed using Fenton’s reaction. The separation was performed by anion-exchange chromatography, extraction chromatography, using TRU and Srresin, and precipitation techniques, followed by α-spectrometry and LSC counting. Tracer solutions and pure ion exchange resins were used to prepare artificial samples and trace nuclides during the analytical procedure. Some real samples of spent ion-exchange resins originating from our TRIGA Mark II research reactor were analyzed.     

Determination of aqueous plutonium oxidation states by solvent extraction

The reliability of two solvent extraction techniques for the determination of Pu oxidation states in solution was tested with low-ionic-strength solutions and with high-Na and high-Mg brines that contained Pu concentrations sufficient for spectrophotometric analysis. One procedure only differentiates between reduced Pu [Pu(III) and Pu(IV)] and oxidized Pu [Pu(V) and Pu(VI)], whereas the second procedure was designed to differentiate between Pu(IV), Pu(V), and Pu(VI) in solution. Both procedures successfully differentiated between oxidized and reduced Pu in both dilute solutions and brines when tested with samples that contained only the Pu(IV), Pu(V), or Pu(VI) oxidation states. However, when the second solvent extraction procedure, which differentiates between Pu(V) and Pu(VI), was employed for solutions that did not contain a strong oxidant to maintain the Pu(VI) oxidation state, significant quantities of Pu(VI) were reduced to Pu(V) during extraction, indicating that accurate quantification of Pu(V) and Pu(VI) is not possible with this procedure.Work supported by the U. S. Department of Energy under Contract DE-ACO6-76RLO 1830.     

Plutonium removal by ion exchange chromatography

A study of Pu recovery at trace level from U solutions by ion exchange technique is presented. Plutonium retention >99.5% onto strong anionic resin, AG-X8, from nitric acid solutions and a 92% recovery using 0.4M HNO₃ at 60°C as eluent, were obtained. Uranium interference in Pu sorption from mixed U/Pu nitrate solutions with low U/Pu ratio (25) was not verified. However, for high U/Pu ratio solutions (10000), uranium interference in Pu retention on the resin, decreases to 59%. Selecting the loading conditions and using AG-X4 resin, 99% Pu retention was achieved. The Pu product is still contaminated with U and another purification cycle is recomended. A scheme for U/Pu first cycle separation is proposed.