Rapid quantification of TBP and TBP degradation product ratios by FTIR-ATR

Tri-n-butyl phosphate (TBP) is the key complexant within the plutonium and uranium reduction extraction process used to extract uranium and plutonium from used nuclear fuel. During reprocessing TBP degrades to dibutyl phosphate (DBP), butyl acid phosphate (MBP), butanol, and phosphoric acid over time. A method for rapidly monitoring TBP degradation is needed for the support of nuclear forensics. Therefore, a Fourier transform infrared spectrometry-attenuated total reflectance (FTIR-ATR) technique was developed to determine approximate peak intensity ratios of TBP and its degradation products. The technique was developed by combining variable concentrations of TBP, DBP, and MBP to simulate TBP degradation. This method is achieved by analyzing selected peak positions and peak intensity ratios of TBP and DBP at different stages of degradation. The developed technique was tested on TBP samples degraded with nitric acid. In mock degradation samples, the 1,235 cm^?1 peak position shifts to 1,220 cm^?1 as the concentration of TBP decreases and DBP increases. Peak intensity ratios of TBP positions at 1,279 and 1,020 cm^?1 relative to DBP positions at 909 and 1,003 cm^?1 demonstrate an increasing trend as the concentration of DBP increases. The same peak intensity ratios were used to analyze DBP relative to MBP whereas a decreasing trend is seen with increasing DBP concentrations. The technique developed from this study may be used as a tool to determine TBP degradation in nuclear reprocessing via a rapid FTIR-ATR measurement without gas chromatography analysis.     

Effect of tri-n-butyl phosphate on physical properties of dodecane–nitric acid system

Extraction power of solvent depends upon the physical properties of the system. Tri-n-butyl phosphate (TBP) in dodecane is a versatile solvent used in the nuclear fuel reprocessing like PUREX process. The study of physical properties like density, viscosity, interfacial tension and solubility for TBP–nitric acid–dodecane system will be helpful in carrying out different extraction studies during PUREX process. Thus, physical properties like density, viscosity, interfacial tension and solubility have been measured for TBP–nitric acid–dodecane system using pycnometer, viscometer, pendant drop method and high performance liquid chromatography respectively. It has been observed that density and viscosity increases but interfacial tension and solubility decreases with the concentration of TBP in dodecane–nitric acid system. Physical properties of 30 % TBP–nitric acid–dodecane system have also been studied in detail. All these studies will also be useful in stripping out dissolved TBP from the nuclear waste.     

Comparative studies on co-extraction of uranium(VI) and different mineral acid from aqueous feed solutions using TBP,TOPO and TOA

Equilibrium and kinetics of co-extraction of hexavalent uranium and mineral acids from aqueous solutions into a hydrocarbon phase (paraffin) using tri n-butyl phosphate (TBP), tri-n-octyl phosphine oxide (TOPO) and tri-n-octyl amine (TOA) has been studied. Relative rates of extraction of uranium(VI) and mineral acid by different complexing ligands were measured simultaneously using bulk-liquid membrane system. Acid extraction by complexing ligands was found to be significant. Wherever there was a possibility of the formation of the third phase, isodecanol was used as an organic phase modifier. Study revealed that isodecanol promotes acid extraction and substantially reduces distribution coefficient of U(VI) into the hydrocarbon phase. The rate of acid extraction by different ligand was in the order of TOPO > TOA > TBP–isodecanol > TBP, whereas the rate of extraction of uranium(VI) was in the order TOPO > TOA > TBP > TBP–isodecanol. A kinetic model was developed to predict concentration of acid and U(VI) in the feed, organic and the strip phase during extraction. The mass transfer coefficients for acid and metal were determined by fitting the model to the observed concentration–time data.     

Properties of PAN-TBP extraction chromatographic material

Three production routes of the preparation of a solid extractant based on tributylphosphate (TBP) embedded in the polyacrylonitrile matrix (PAN) have been studied. The method of direct PAN coagulation with TBP was found to be not viable due to the significant TBP solubility in the coagulation bath. The most suitable PAN-TBP solid extractant was prepared by the well-known impregnation method of ready-made neat PAN beads. The kinetics of uranium extraction from 3 mol L^?1 HNO₃, the effect of nitrate and nitric acids concentrations on the value of weight distribution coefficients D _g as well as the uranium “extraction isotherm” were determined for this material. Uranium extraction was rather fast, approximately 1 h was sufficient for the equilibrium achievement. Capacity for the uranium uptake, measured in batch experiments on PAN-TBP for 0.048 mol L^?1 of uranium in 3 mol L^?1 nitric acid, was found to be q = 0.363 mmol g^?1 (58 % of the theoretical capacity). It was concluded that PAN-TBP material behaves like TBP in liquid–liquid extraction. Extraction capacity determined in column experiments was lower (by about 23 %) than expected from the “extraction isotherm” due to the TBP leaching out of the column. The thus prepared material is therefore not very suitable for multicycle extraction and stripping and can be used once, particularly for the analytical purposes.     

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.     

Quantitative determination of tri-n-butyl phosphate by fluorescent X-ray spectrometry through inorganic association with an actinide

Tri-n-butyl phosphate (TBP) continues to be the most widely used solvent in nuclear fuel extraction, refining and reprocessing units for the extraction of actinides and their separation from fission products. An X-ray fluorescence spectrometric method (XRFS) for the determination of TBP content with an X-ray detectable element is presented. The method involves formation of an ion association complex of uranium with TBP-kerosene mixture in 3M nitric acid. The analytes uranium and bromine used as internal ratio elements in organic extract are excited by a primary X-ray beam from a rhodium tube. The solvent concentration is determined from the ratioed characteristic intensities of uranium and bromine. The procedure permits the determination of organic solvent in the range 0.5 to 5.0% with a relative standard deviation of 0.1%.     

Synergistic extraction of uranium with mixtures of PC88A and neutral oxodonors

The extraction of uranium(VI) from nitric acid medium is investigated using 2-ethylhexyl phosphonic acid-mono-2-ethylhexyl ester (PC88A in dimeric form, H₂A₂) as extractant either alone or in combination with neutral extractants such as tri-n-butyl phosphate (TBP), trioctyl phosphine oxide (TOPO), and dioctyl sulfoxide (DOSO). The effects of different experimental parameters such as aqueous phase acidity (up to 10 M HNO₃), nature of diluent [xylene, carbon tetrachloride (CCl₄), n-dodecane and methyl iso-butyl ketone (MIBK)] and of temperature (303–333 K) on the extraction behavior of uranium were investigated. Synergistic extraction of uranium was observed between 0.5 and 6 M HNO₃. Use of MIBK as diluent was also studied. Temperature variation studies using PC88A as extractant showed exothermic nature of extraction process. Studies were carried out to optimize the conditions for the recovery of uranium from the raffinate generated during the purification of uranium from nitric acid medium. Inductively Couple Plasma Atomic Emission Spectroscopy (ICP-AES) and Energy Dispersive X-Ray Fluorescence (EDXRF) techniques were employed for analysis of uranium in equilibrated samples.     

Improved determination of tributyl phosphate degradation products (mono- and dibutyl phosphates) by ion chromatography

Tributyl phosphate (TBP) is a very important compound in the nuclear industry, particularly in the area of nuclear fuel reprocessing. This compound is used in the PUREX (plutonium and uranium refining extraction) process which consists of the extraction of uranium and plutonium from an aqueous nitric acid phase, for the purpose of recycling. But TBP may be degraded to dibutyl phosphate (DBP) and monobutyl phosphate (MBP) by dealkylation of one or two butoxy groups, respectively. We have compared and evaluated the capacity of two resins manufactured by Dionex (AS11 and AS5A) in the separation and measurement of these two degradation products. AS11 generates two interferences: nitrite/DBP and carbonate/MBP. The first one is the most serious. So, we have developed a method for oxidising nitrite ions to nitrate ions which have no trouble over the measurement. The second resin tested, AS5A, allows a very efficient separation between DBP and NO₂⁻ ions and a good separation between MBP and CO₃²⁻ in comparison with the AS11. The detection limits for the AS5A column are 0.13 μM for MBP and 0.71 μM for DBP (injection LOOP=50 μl).     

Microwave-assisted dissolution of ceramic uranium dioxide in TBP–HNO3 complex

Microwave-assisted dissolution of ceramic uranium dioxide in tri-n-butyl phosphate (TBP)–HNO₃ complex was investigated. The research on dissolution of ceramic uranium dioxide in TBP–HNO₃ inclusion complex under microwave heating showed the efficiency of the use of this method. Nitric acid present in the inclusion complex participates both dissolution of UO₂, and oxidation of U(IV)–U(VI), the resulting UO₂(NO₃)₂ extracted with tri-n-butyl phosphate. Dissolution rate depends on both temperature of microwave dissolution process, and concentration of nitric acid present in the inclusion complex. The most intensive dissolution process is when the concentration of nitric acid ≥2 mol/L and the temperature of 120 °C. From the experimental data obtained by two kinetic models activation energies were calculated. At the average activation energy of UO₂ dissolution in TBP–HNO₃ complex equal 70 kJ/mol, and reaction order is close to one, i.e. the reaction takes place in an area close to kinetic.     

Extraction of uranium(VI) from nitric acid medium by 1.1M tri-n-butylphosphate in ionic liquid diluent

Summary A systematic study on the extraction of U(VI) from nitric acid medium by tri-n-butylphosphate (TBP) dissolved in a non-traditional diluent namely 1-butyl-3-methylimidazolium hexafluorophosphate (bmimPF₆) ionic liquid (IL) is reported. The results are compared with those obtained using TBP/n-dodecane (DD). The distribution ratio for the extraction of U(VI) from nitric acid by 1.1M TBP/bmimPF₆ increases with increasing nitric acid concentration. The U(VI) distribution ratios are comparable in the nitric acid concentration range of 0.01M to 4M, to the ratios measured using 1.1M TBP/DD. In contrast to the extraction behavior of TBP/DD, the D values continued to increase with the increase in the concentration of nitric acid above 4.0M. The stoichiometry of uranyl solvate extracted by 1.1M TBP/IL is similar to that of TBP/DD system, wherein two molecules of TBP are associated with one molecule of uranyl nitrate in the organic phase. Ionic liquid alone also extracts uranium from nitric acid, albeit to a small extent. The exothermic enthalpy accompanying the extraction of U(VI) in TBP/bmimPF₆ decreases with increasing nitric acid and with TBP concentrations.     

Counter-current extraction and separation of U(VI) from a mixture of U(VI)–Th(IV)–Y(III) using tris-2-ethyl hexyl phosphate (TEHP)

Evaluation of tris-2-ethyl hexyl phosphate (TEHP) for counter-current extraction and separation of U(VI) from a mixture of U(VI)–Th(IV)–Y(III) from nitric acid medium was carried out under wide experimental conditions. Batch extraction studies were carried out to investigate the effect of nitric acid concentration in feed solution, U(VI)/Th(IV) ratio and extractant concentration and the results were compared with established solvent such as tri-n-butyl phosphate (TBP) for separation of U(VI) from nitric acid medium. McCabe–Thiele diagrams for extraction as well as stripping of U(VI) were constructed under simulated conditions. Based on batch experiments, six stage counter-current extraction studies were conducted under various TEHP concentration and it was observed that 0.1 M TEHP/n-paraffin was most suitable for selective recovery of U(VI) from a mixture of U(VI)–Th(IV). An optimized condition, 0.1 M TEHP/n-paraffin, 2 M HNO₃ in feed and six number of stages was evaluated for selective extraction and stripping of U(VI) from a solution containing mixture of U(VI)–Th(IV)–Y(III) in nitric acid medium. The U(VI) in strip solution was precipitated using 30 % H₂O₂ at pH ~3. Average particle size of the final precipitate was found to be ~33 μm.     

Hexanoic acid as an alternative diluent in a GANEX process: feasibility study

Used nuclear fuel is radiotoxic for mankind and its environment for a long time. However, if it can be transmuted, the radiotoxicity as well as its heat load are reduced. Before a transmutation the actinides within the used fuel need to be separated from the fission, corrosion and activation products. This separation can be achieved by using the liquid–liquid extraction technique. One extraction process that can be used for such a separation is the Group ActiNide EXtraction (GANEX) process. One GANEX process that can successfully accomplish the separation utilizes the diluent cyclohexanone in combination with the extractant tributylphosphate (TBP) (30 % vol) and a second extractant, CyMe₄-BTBP (10 mM). However, there are some issues when using cyclohexanone as diluent. In this work an alternative diluent has therefore been tried in order to determine if it can replace cyclohexanone. The diluent used was hexanoic acid. In a system containing 10–12 mM CyMe₄-BTBP and 30 % vol TBP in hexanoic acid with the aqueous phase 4 M HNO₃, the distribution ratios for americium and curium are unfortunately low (D _Am = 1.1 ± 0.27, D _Cm = 1.6 ± 1.81). The concentration of CyMe₄-BTBP ligand, the extractant of curium and americium, could unfortunately not be increased, because of limited solubility in hexanoic acid. The distribution ratios for fission, corrosion and activation products were low for most metals; however, cadmium, palladium and molybdenum all unfortunately have distributions ratios above 1. To conclude, low americium and curium extractions indicate that hexanoic acid is not a suitable diluent which could replace cyclohexanone in a GANEX process.     

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.     

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.     

Wet-oxidation of spent organic waste tributyl phosphate/diluents

Tributyl phosphate was used in reprocessing of spent nuclear fuel inthe Purex process. The amount of uranium retained in the organic phase dependson the type of TBP/diluent. Destruction of spent TBP is of high interest inwaste management. The use of the oxidative degradation of TBP diluted withkerosene, carbon tetrachloride, benzene and toluene using potassium permanganateas oxidant was carried out to produce stable inorganic dry particle residuewhich is then immobilized in different matrices. The different factors affectingthe destruction of spent waste were investigated. The uptake and decontaminationfactor for both ^{152, 154}Eu and ¹⁸¹Hf and the analysisof the final product have been studied.     

Separation of uranium and thorium using tris(2-ethylhexyl) phosphate as extractant

The distribution behavior of uranium and thorium has been investigated in a biphasic system of different aqueous nitric acid concentrations and a solution of tris(2-ethylhexyl) phosphate (TEHP) inn-dodecane at 25°C. The effect of different uranium and thorium concentrations in the aqueous phase on the extraction of these metal ions is evaluated. These results indicate that TEHP is a better choice than tri-n-butyl phosphate (TBP) for the separation of²³³U from the irradiated thorium matrix.     

Preparation of125I labeled testosterone derivatives for use in radioimmunoassays

A study for separation and sequential recovery of uranium and plutonium from nitric acid solutions by extraction chromatography using tributyl phosphate (TBP)/Amberlite XAD7 as stationary phase is presented. Distribution ratios of actinides, lanthanides and fission products were obtained. The column capacity was investigated and actinides retention conditions were established. Finally, U-Pu sequential separation was studied as well as the U and Pu recovery yields from nitric solutions containing Am/fission products were determined.     

Synergistic extraction of uranium(VI) with tri-n-butyl phosphate and i-butyldodecylsulfoxide

Summary The synergistic extraction of uranium(VI) from aqueous nitric acid solution with a mixture of tri-n-butyl phosphate (TBP) and i-butyldodecylsulfoxide (BDSO) in toluene was investigated. The effects of the concentrations of extractant, nitric acid, sodium nitrate and sodium oxalate on the distribution ratios of uranium(VI) have been studied. The values of enthalpy change for the extraction reactions with BDSO, TBP and a mixture of TBP and BDSO in toluene were -23.2±0.8 kJ/mol, -29.2±1.4 kJ/mol and -30.6±0.6 kJ/mol, respectively. It has been found that the maximum synergistic extraction effect occurs when the molar ratio of TBP to BDSO is close to 1. The composition of the complex of the synergistic extraction is UO₂(NO₃)₂ ^. BDSO ^. TBP.     

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.     

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium in fast breeder reactors (FBRs) fuel reprocessing

Ammonium uranyl carbonate (AUC) based process of simultaneous partitioning and reconversion for uranium and plutonium is developed for the recovery of uranium and plutonium present in spent fuel of fast breeder reactors (FBRs). Effect of pH on the solubility of carbonates of uranium and plutonium in ammonium carbonate medium is studied. Effect of mole ratios of uranium and plutonium as a function of uranium and plutonium concentration at pH 8.0–8.5 for effective separation of uranium and plutonium to each other is studied. Feasibility of reconversion of plutonium in carbonate medium is also studied. The studies indicate that uranium is selectively precipitated as AUC at pH 8.0–8.5 by adding ammonium carbonate solution leaving plutonium in the filtrate. Plutonium in the filtrate after acidified with concentrated nitric acid could also be precipitated as carbonate at pH 6.5–7.0 by adding ammonium carbonate solution. A flow sheet is proposed and evaluated for partitioning and reconversion of uranium and plutonium simultaneously in the FBR fuel reprocessing.