EXTRACTION OF PLUTONIUM (III) BY TBP IN PRESENCE OF URANIUM

ABSTRACT

The distribution ratio (D) values for the extraction of plutonium (III) from nitric acid medium into 30% TBP in n-dodecane saturated with uranium(VI) (0% to 80%) were determined. For a fixed saturation of TBP with uranium, the D values for Pu(III) were found to increase with increase in nitric acid concentration (1M to 5M). At a fixed nitric acid concentration, the D values were found to decrease with increase in loading of TBP with uranium. The D values for the extraction of Pu(III) using 20% TBP in n-dodecane and 30% TBP in n-paraffin at 80% uranium saturation were also determined The distribution data was least squares analysed against concentration of HNO₃ as well as percentage saturation of TBP with uranium and the coefficients obtained are reported. For all these extraction systems, D values for U(VI) were also determined.     

COEXTRACTION OF URANIUM AND TECHNETIUM IN TBP-SYSTEMS

Abstract

The coextraction of technetium with uranium in TBP/HNO systems has been studied. The extraction mechanism,

has been found to agree well with experimental data at 5 M HNO,.The extraction constant, K_TC,e for this reaction Is given by

where T is temperature (K) and C_T total TBP concentration (%). This equation is valid for 5 M HNO, DSFS–30% TBP, and from 20 to 60°C, with an average error of 2·8% in log K_TC,e     

Interpretation of the Extraction Mechanism of the Purex and Thorex Processes from Kinetics Data

Abstract

The rate of transfer of species between phases in solvent extraction is expected to be controlled by chemical reaction kinetics, diffusion, or a combination of these steps. Transfer kinetics data for uranium(VI), thorium(IV), and zirconium(IV) indicate that the chemical reaction is the rate-controlling step for most dispersed systems. The evidence for this conclusion is: (1) the transfer rate constants for the extraction of uranium were the same for Lewis cell, single drop, and Kenics mixer tests; (2) the forward transfer rate constants increase with increasing TBP concentration (the viscosity of the organic phase also increases) in all cases; (3) the reverse transfer rate varies dramatically with changes in nitrate concentration in the aqueous phase (same phase viscosities); and (4) the difference in the activation energies of the forward and reverse reactions for uranium(VI) transfer is equal to the heat of reaction of uranium with TBP. A proposed mechanism for the solvent extraction kinetics which is consistent with the experimental results includes the formation or decomposition of a charged interfacial complex as the rate-controlling step and equilibration of the interfacial complex with the bulk organic complex as the rapid step.     

Batch Tests for Optimisation of Solvent Composition and Process Flexibility of the CHALMEX FS-13 Process

ABSTRACT

Studies have been performed with the purpose of determining the optimal solvent composition of a Chalmers grouped actinide extraction (CHALMEX) solvent for the selective co-extraction of transuranic elements in a novel Grouped ActiNide EXtraction (GANEX) process. The solvent is composed of 6,6’-bis(5,5,8,8-tetramethyl-5,6,7,8-tetrahydro-benzo-[1,2,4]-triazin-3-yl)-[2,2’]-bipyridine (CyMe₄-BTBP) and tri-n-butyl phosphate (TBP) in phenyl trifluoromethyl sulfone (FS-13). The performance of the system has been shown to significantly depend on the ratios of the two extracting agents and the diluent to one another. Furthermore, the performance of the determined optimal solvent (10 mM CyMe₄-BTBP in 30% v/v TBP and 70% v/v FS-13) on various simulated PUREX raffinate solutions was tested. It was found that the solvent extracts all transuranic elements with high efficiency and good selectivity with regard to most other elements (fission products/activation products) present in the simulated PUREX raffinate solutions. Moreover, the solvent was found to extract a significant amount of acid. Palladium, silver, and cadmium were co-extracted along with the TRU-radionuclides, which has also been observed in other similar CHALMEX systems. The extraction of plutonium and uranium was preserved for all tested simulated PUREX raffinate solutions compared to experiments using trace amounts.     

Separation and Recovery of Uranium,Neptunium, and Plutonium from High Level Waste Using Tributyl Phosphate: Countercurrent Studies with Simulated Waste Solution

ABSTRACT

The present work deals with countercurrent extraction studies on the partitioning of uranium, neptunium, and plutonium using 30% tributyl phosphate (TBP) from simulated high level waste solution generated during reprocessing of spent uranium fuel from pressurized heavy water reactors. The oxidation states of neptunium and plutonium were adjusted either by 0.01 M potassium dichromate or 0.01 M dioxovanadium ion. Neptunium and plutonium, extracted in the TBP phase, were stripped together using a mixture containing 0.05 M ascorbic acid and 0.25 M hydrogen peroxide in 2.0 M nitric acid solution. Although dioxovanadium ion is more effective for proper adjustment of the oxidation states of plutonium and neptunium, subsequent recovery of these actinides from loaded TBP is better if potassium dichromate is used for the valency adjustment. Results of the stagewise analysis of extraction and stripping of actinides using mixer-settlers are presented.     

DISTRIBUTION OF N- AND ISO- BUTYRALDEHYDES BETWEEN TRI-N-BUTYL PHOSPHATE/ N-DODECANE AND NITRIC ACID

ABSTRACT

The distribution of n- and iso-butyraldehydes between tri-n-butyl phosphate(TBP) n-dodecane(nDD) and HNO₃ were measured. The distribution ratio of n-butyraldehyde in the TBP/nDD and HNO₃ system was nearly the same as that of iso-butyraldehyde. The distribution ratios of n- and iso-butyraldehydes increased with TBP concentration in the organic phase. The equilibrium constant of the extraction reaction was about 2. In a uranium, neptunium and plutonium separation process, most of the n- and iso-butyra1dehydes fed into theNp separation stepor into thePu/U partition will be left with the TBP solvent. The two compounds will be partly back-extracted to the aqueous phase in the U purification and in the solvent washing steps of the PUREX process.     

Synergistic Effects in the Extraction of Uranium(VI) by Di-4-octylphenyl Phosphoric Acid

Abstract

The extraction of uranium(VI) from sulfuric acid solutions by di-4-octylphenyl phosphoric acid (DOPPA) is enhanced by the addition of neutral organophosphorus compounds due to synergistic action. The effect of tri-n-butyl phosphate (TBP), dibutylbutyl phosphonate (DBBP), and tri-n-octyl phosphine oxide (TOPO) was studied. The synergistic effect increased in this order. In the case of TBP and DBBP the extraction coefficient for U(VI) decreased with increasing concentration of synergistic agent after reaching a maximum. With TOPO, on the other hand, there was an increase even after this limit. This was because of the extraction of uranium by TOPO itself. The effect of uranium loading in the organic phase on the synergistic behavior was studied and the results were compared with those obtained with di-2-ethylhexyl phosphoric acid (DEHPA) in the presence of the same synergistic agents. The results with these two extractants indicate that with TOPO the synergism is mainly due to the formation of substitution products of the type UO₂A₂B₂ and with TBP addition products of the type UO₂(HA₂)₂B.     

Recent R&D Studies Related to Coprocessing of Spent Nuclear Fuel Using N,N-Dihexyloctanamide

Abstract

The PUREX process has undergone several modifications to address the issues of high burn up, fewer solvent extraction cycles, and reduced waste arisings. Advanced fuel cycle scenarios have led to a renewed international interest in the development of separation schemes for co-recovering U/Pu from spent fuels. Completely incinerable N,N-dihexyloctanamide (DHOA) has been identified as a promising candidate for the reprocessing of spent fuels. Batch extraction studies were carried out to evaluate DHOA and TBP for the coprocessing (co-extraction and co-stripping) of U and Pu from spent fuel under varying concentrations of nitric acid and of uranium as well as under simulated pressurized heavy water reactor spent fuel feed conditions. At 50 g/L U in 4 M HNO₃, D_Pu values for 1.1 M DHOA and 1.1 M TBP solutions in n-dodecane were 7.9 and 3.8, respectively. In contrast, significantly lower D_Pu value at 0.5 M HNO₃ (4 × 10^?3) for DHOA as compared to TBP (4 × 10^?2) suggested that it was a better choice for coprocessing of spent nuclear fuel. This behavior was attributed to the change in stoichiometry of extracted species at lower acidity vis-a-vis the higher acidity. These studies suggest that plutonium fraction can be enriched with respect to uranium contamination in the product stream. DHOA displays better extraction behavior of plutonium and stripping behavior of uranium under simulated feed conditions. DHOA appears distinctly better than TBP with respect to fission product/structural material decontamination of U/Pu.     

DISTRIBUTION OF NITROUS ACID BETWEEN TRI-N-BUTYL PHOSPHATE/ N-DODECANE AND NITRIC ACID

ABSTRACT

The distribution of nitrous add between tri-n-butyl phosphate ( TBP) n-dodecane ( nDD) and nitric add has been measured as functions of nitrous and nitric add concentrations in the aqueous phase and functions of TBP and uranium concentrations in the organic phase

The extraction of nitrous add by TBP can be described by the following reaction:     

THIRD PHASE FORMATION IN NITRIC ACID EXTRACTION BYn-OCTYL(PHENYL)-N,N-DIISOBUTYLCARBAMOYL-METHYLPHOSPHINE OXIDE

ABSTRACT

The third phase formation was studied as a function of n-octyl(phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tri-M-butyl phosphate (TBP) concentrations and temperature in the extraction of nitric acid. The concentration fractions of CMPO and TBP in the second and the third phases were determined by gas chromatography. Both CMPO and TBP were found to be enriched in the third phase. The concentrations of nitric acid in the second and the third phases relatively agreed with the calculated concentrations based on the extraction equilibrium constants of nitric acid by CMPO and TBP. The extraction of Np with the third phase formation was also discussed.     

Solvent Extraction Chemistry and Kinetics of Zirconium

Abstract

The solvent extraction behavior of zirconium in the HN03-tributyl phosphate (TBP) system can be explained based on the existence of four principal aqueous species, Zr⁴⁺, ZrOH³⁺, Zr₃(OH)⁸⁺ ₄, and oxo-polymers. The Zr⁴⁺ and ZrOH³⁺ species are extractable and are in equilibrium with inextractable Zr₃(OH)⁸⁺ ₄. The oxo-polymers are formed by heat, are-inextractable, and are not: in equilibrium with the other species. The aqueous equilibria and their equilibrium quotients have been previously determined. In the present study, these equilibria were used along with both tracer and macro zirconium concentrations (oxo-polymers excluded by extraction and back scrubbing) to determine the distribution equilibrium constants for both the Zr⁴⁺ and ZrOH³⁺ ions. The four equilibrium constants give excellent fits to both tracer and macro-zirconium distribution data.

The concentrations of the extractable zirconium species which are calculated from the equilibria have been used to begin examining the extraction kinetics of zirconium in the HNO₃-TBP system. In relatively concentrated nitric acid, approximately 3 M and greater, Zr⁴⁺ ion predominates, and the rate of extraction of zirconium increases as approximately the second power of the TBP concentration. In low acid (1 M and less) ZrOH³⁺ ion predominates, and the rate of extraction of zirconium increases as approximately the third power of the nitrate concentration. This is in significant contrast with the behavior of uranium, which shows only a small dependence of the extraction rate on TBP concentration, and no dependence on nitrate concentration. This suggests that operation of a kinetic separations system at low TBP and nitrate concentrations will significantly improve separations over those achieved at equilibrium.     

Studies on the Development of a Flow-Sheet for AHWR Spent Fuel Reprocessing Using TBP

The present paper describes the results of solvent extraction studies carried out in batch mode to collect data on distribution of uranium, plutonium, and thorium using 5% TBP in n-dodecane. Extraction studies are carried out from feed solutions having bulk thorium containing aluminum and fluoride ions in ～3.00–4.00 M nitric acid at concentration levels anticipated in feed solutions during Advanced Heavy Water Reactor (AHWR) spent fuel reprocessing. Studies are carried out under varied experimental conditions. Parameters such as organic to aqueous phase ratio during extraction, concentration of nitric acid for scrubbing co-extracted thorium from loaded organic phase etc., are studied in detail. Hydroxylamine nitrate is selected for reductive stripping of plutonium in preliminary studies. Reagent mixture containing 0.30 M HAN + 0.60 M HNO₃ and 0.20 M N₂H₄ is found to be optimum for plutonium partitioning. This paper also describes the extraction and stripping of uranium and plutonium in co-current mode. The extraction behavior of relevant fission products is studied from a simulated feed solution. A preliminary study on a few commercially available reducing agents is also included. These data are useful in developing a flow-scheme for the recovery of uranium and plutonium from spent fuel originating from AHWR.     

Note: Effect of Temperature in the Extraction of Uranium and Plutonium by Triisoamyl Phosphate

Abstract

The extraction of uranium(VI) by triisoamyl phosphate (TiAP) has been studied to derive the thermodynamic parameters such as entropy change and the free-energy change. The extraction of U(VI) and Pu(IV) has also been studied with 1.1 M solutions of tri-n-butyl phosphate (TBP), tri-n-amyl phosphate (TAP), and TiAP as a function of temperature. While the enthalpy of U(VI) extraction was found to be exothermic, the enthalpy for the extraction of Pu(IV) was always found to be endothermic. The temperature at which the distribution ratios of U(VI) and Pu(IV) cross each other (the temperature of inversion) has been derived for TBP, TAP, and TiAP, and the results reveal the lowest temperature of inversion occurs for TiAP. The results indicate the advantage of TiAP as an extractant in avoiding plutonium reflux during the PUREX process involving high plutonium feed solutions, in addition to lower aqueous solubility, freedom from the third-phase formation problem, lower degradation, and better economics.     

Separation of radioactive metal ions by hollow fiber-supported liquid membrane and permeability analysis

Extraction and stripping of uranium ions from nitrate media using a hollow fiber liquid membrane contactor was studied. In this study, tri-butyl phosphate (TBP) diluted in kerosene was used as extractant and sodium hydroxide was applied as a stripping solution. Uranium ions were extracted using TBP 5%(v/v) by rejecting thorium ions into raffinate with a maximum percentage of extraction for the uranium being 67%. The mathematical model was focused on the extraction side of the liquid membrane system. The permeability for each concentration of HNO₃ was investigated by mass transfer theory. When the concentration of HNO₃ increased, more uranium ions were extracted; however, when the concentration of uranium and thorium in the feed solution was increased, the percentage of extraction and stripping slightly decreased because the permeability decreases when the concentration of the feed solution increases due to membrane fouling and concentration polarization.     

OSCILLATORY EXTRACTION OF URANIUM

ABSTRACT

The kinetics of U and Ce extraction by tri-n-butyl-phosphate ( TBP) and the redox potential behavior in the aqueous phase have been investigated in an oscillatory extraction system. The extraction has been followed continuously by UV-visible spectroscopy, potentiometry, and radioactive counting methods. Data collected showed the changes of Ce⁴⁺ and U0²⁺ concentrations in organic and aqueous phases with time. Uranium actively responded to the redox potential fluctuations. The uranium distribution coefficient changed repeatedly and was out-of-phase with Ce( IV) fluctuations.     

Pseudo-Emulsion Based Hollow Fiber Strip Dispersion Technique (PEHFSD): Optimization,Modelling and Application of PEHFSD for Recovery of U(VI) from Process Effluent

Abstract

Pseudo emulsion based hollow fiber strip dispersion technique (PEHFSD) is the first of its kind ever explored in radioactive environment for the extraction of uranium from acidic process streams. Permeation of U(VI) was investigated as a function of various experimental variables such as hydrodynamic conditions (flow rates of pseudo-emulsion and feed phase), concentration of U(VI) in the feed phase, concentration of tri-n-butylphosphate (TBP), HNO₃ concentration in feed phase, O/A ratio and 0.01 M HNO₃ as stripping agent in pseudo-emulsion phase. The mass transfer coefficient was calculated from the experimental results and a model has been presented for determining mass transfer characteristics. PEHFSD has been demonstrated for separation/recovery of uranium from oxalate supernatant waste generated during plutonium precipitation by oxalic acid. PEHFSD and HFSLM (hollow fiber supported liquid membrane) performance has been compared in order to analyze the efficiency of the technique.     

Application of Extraction Chromatography to Nuclear Fuel Reprocessing

Abstract

The potentialities of applying extraction chromatography to the reprocessing of reactor fuels on an industrial scale have been investigated. The stationary phase was undiluted (100%) tri-n-butyl phosphate (TBP) and the mobile phases were nitric acid or nitrate salt solutions with or without reducing agents for plutonium.

Several extraction chromatographic processes for the recovery of nuclear grade uranium and plutonium are described. The flowsheets are based on a systematic determination of the distribution coefficients of relevant metal species (particularly those of uranium, neptunium, plutonium, americium, ruthenium, zirconium and niobium) in the chromatographic systems employed.

The Purochromex process developed for the recovery of uranium and plutonium from light-water reactor fuels and the Eurochromex process developed for the separation of highly enriched uranium from irradiated U/A1 alloy, U/Zr alloy and uranyl sulfate fuels have successfully been hot-tested on a laboratory scale and cold-tested on an “industrial scale.”

Some complementary studies related to the separation processes. such as radiation degradation of the stationary phase and the removal of tributyl phosphate from product and waste streams, are also described.     

Role of Acetohydroxamic Acid in Selective Extraction of Technetium and Uranium Employing N,N-Dihexyloctanamide as Extractant

Straight chain N,N-dihexyloctanamide (DHOA) has been identified as a promising alternate extractant to tributyl phosphate (TBP) for the reprocessing of uranium based spent fuels. The present work compares extraction behavior of technetium using DHOA and TBP solutions in n-dodecane, under varying experimental conditions such as acidity (0.5–6 M HNO₃); extractant concentration (1.1 and 1.5 M), and uranium loading (50 g/L, relevant for Pu rich spent fuel feed solutions). The effect of acetohydroxamic acid concentration on U, Pu, Np, and Tc extraction behavior has also been investigated. Pu(IV)-AHA interaction and its influence on extraction using TBP and DHOA extractants has been studied spectrophotometrically. The experimental data suggest that 1.1 M DHOA is better than 1.1 M TBP with respect to co-extraction of Tc and U, and U decontamination with respect to Np/Pu.     

Modeling of Liquid-Liquid Extraction (LLE) Equilibria Using Gibbs Energy Minimization (GEM) for the System TBP–HNO3–UO2–H2O–Diluent

Liquid-liquid extraction (LLE) is a widely used separation method for an extensive range of metals including actinides. The Gibbs energy minimization (GEM) method is used to compute the complex chemical equilibria for the LLE system HNO₃–H₂O–UO₂(NO₃)₂–TBP plus diluent at 25°C. The nonelectrolyte phase is treated as an ideal mixture defined by eight tri-n-butyl phosphate (TBP) complexes plus the inert diluent. The Pitzer method is used to capture nonidealities in the concentrated electrolyte phase. The generated extraction isotherms are in very good agreement with reported experimental data for various TBP loadings and electrolyte concentrations demonstrating the adequacy of this approach to analyze complex multiphase multicomponent systems. The model is robust and yet flexible allowing for expansion to other LLE systems and coupling with computational tools for parameter analysis and optimization.     

Solvent Extraction and Spectrophotometric Studies on Synergistic Extraction of Plutonium(IV) by Mixtures of Thenoyltrifluoroacetone (HTTA) and Tri-n-butylphosphate (TBP) in Benzene

Abstract

The synergistic extraction of Pu(IV) from perchloric acid solutions into mixtures of thenoyltrifluoroacetone (HTTA) and tri-n-butylphosphate (TBP) in benzene was investigated by solvent extraction methods. The adduct responsible for synergism was found to be Pu(TTA)₄·TBP. The adduct formation between Pu(TTA)₄ and TBP in the benzene phase was also investigated by spectrophotometry. The equilibrium constants for the equilibria involved were obtained both by solvent extraction and by spectrophotometric methods.