Electro-oxidative process for the mineralization of solvent and recovery of plutonium from organic wastes

Mediated electrochemical oxidation is a promising technique for the destruction of organic compounds. Destruction of tributyl phosphate (TBP) in normal paraffin hydrocarbon (NPH) in nitric acid medium containing electro-generated Ag(II) was studied. Initially, the effect of uranium, presence of DBP along with uranium in the organic phase and direct electrochemical oxidation without catalyst (Ag) on the destruction of 30% TBP/NPH system was evaluated. For a comparison, the rate of destruction of NPH alone was studied. Further, radioactive laboratory waste solution was tested for the destruction of organic waste under similar experimental conditions. The electrolyte used in the system was 0.5 M AgNO₃ in 8 M HNO₃ at 333 K. The uniqueness in all these experiments is the use of a double end open porcelain diaphragm for the isolation of electrodes. Though there would be a slight reduction in the efficiency, two major hurdles viz., reduction in the concentration of nitric acid and reduction in the volume of catholyte resulting in an increase in cell voltage were avoided. The problem of the migration of Ag⁺/Ag²⁺ and accumulating at the cathode site was overcome by using double end open diaphragm and thorough mixing. The results revealed that the rate of destruction of organics is favoured in the presence of uranium in organic phase and with increase in temperature.     

Uphill transport of uranium across a liquid membrane

A liquid membrane was prepared by entrapping tributyl phosphate (TBP) in a cellulose triacetate (TAC) membrane matrix. The membrane was used to separate two aqueous solutions, one acidic and the other alkaline, which were saturated with TBP to prevent its loss from the membrane. Uphill transport of uranium was achieved with the TBP liquid membrane. Both solutions containing TBP were stirred magnetically. When the initial concentration of uranium in the two solutions was 3.5 mM, more than 50% of the uranium contained in the acidic solution was transported to the alkaline solution across the liquid membrane within 5 h. A transport mechanism is described in which the membrane-bound TBP acts as a mobile carrier for uranium.     

Evaluation of cerium–zirconium mixed oxides for separation of 125Sb from radioactive liquid waste

Journal of Radioanalytical and Nuclear Chemistry - Low level radioactive liquid waste (LLW) contains various radioisotopes like 125Sb, 106Ru, 99Tc and traces of 137Cs, 134Cs, 90Sr. Chemical...     

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.     

Recent R&D towards Aqueous Reprocessing of FBR Fuels

The mixed Pu-rich carbide spent fuel with a burn up of 155 GWd/t from the Fast Breeder Test Reactor is being reprocessed in a hot-cell facility by PUREX process. Based on the input from the operation of this facility, R&D activities were carried out to improve the recovery, decontamination factors, economy and to reduce the waste volumes. Reduction of uranyl ions in a continuous flow electrochemical reactor and electrolytic as well as chemical reduction of 4M HNO₃ from liquid waste could be accomplished in continuous mode. Using the optimized parameters, suitable electrolytic cells/experimental setups were designed for the plant capacity of 6 L/h. Studies on the extraction kinetics of Ru with 30% TBP in NPH revealed that better decontamination factor with respect to Ru can be achieved using fast contactors like centrifugal extractors (CEs). Towards developing a spent solvent recovery system to reduce organic waste volumes, a pilot plant was set up, which could recover diluent as top product of distillation column and 40% TBP as bottom product from inactive degraded solvent. A solvent recovery system using short path distillation was also developed for installation in hot cells.     

A novel rapid detection approach for the analysis of radionuclides in environmental samples using graphite MALDI mass spectrometry

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.     

Method of separating uranium from iron and thorium

Acid leaching of uranium deposits is not a selective process. Sulfuric acid solubilizes iron(III) and half or more of the thorium depending on the mineralog of this element. In uranium recovery by solvent extraction process, uranium is separated from iron by an organic phase consisting of 10 vol% tributylphosphate(TBP) in kerosine diluent. Provided that the aqueous phase is saturated with ammonium nitrate or made 4–5 M in nitric acid prior to extraction. Nitric acid or ammonium nitrate is added to the leach solution in order to obtain a uranyl nitrate product. Leach solutions containing thorium(IV) besides iron are treated in an analogous fashion. Uranium can be extracted away from thorium using 10 vol% TBP in kerosine diluent. The aqueous phase should be saturated with ammonium nitrate and the pH of the solution lowered to 0.5 with sufficient amount of sulfuric acid. In other words, the separation of uranium and thorium depends on the way the relative distributions of the two materials between aqueous solutions and TBP vary with sulfuric acid concentration. Thorium is later recovered from the waste leach liquor, after removal of sulfate ions. Uranium can be stripped from the organic phase by distilled water, and precipitated as ammonium diuranate.     

Radiochemical analysis of uranium isotopes in soil and sediment samples with extraction chromatography

An accurate and simple analytical technique for uranium isotopes in highly contaminated soil samples was developed and validated by application to IAEA-Reference samples and environmental samples. For overcoming the demerits of the TBP extraction method, sample materials were decomposited with HNO(3) and HF and uranium isotopes were purified with an anion exchange resin and a TRU Spec resin. With the extraction chromatography method, hindrance elements were completely removed from the uranium fraction. The chemical yields with the extraction chromatography method were <10% higher than those with the TBP extraction method. The concentrations of uranium isotopes using the extraction chromatography method were consistent with the reference values reported by the IAEA.     

Microtitrimetric determination of uranium in tributylphosphate (30%)-n-dodecane-nitric acid medium in the presence of excess of iron, magnesium and aluminium

A method is described for microdetermination of uranium in aqueous phases containing nitric acid, iron, magnesium, aluminium and traces of TBP, and in organic phases containing TBP, n-dodecane and nitric acid. Treatment of the organic phase strips the uranium, reduces it to U(IV), and destroys any nitrite present.     

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.     

Radioactivity measurement with a plastic scintillation vial

Liquid scintillation counting is widely used to measure radioactivity, but it generates radioactive organic liquid waste. Not to generate the liquid waste using a liquid scintillation counter, novel counting method with a plastic scintillation vial was designed. The counting efficiency for ³²P was 10–40% and that for ¹²⁵I was 4–8%. The efficiency depended on the sample volume. The color quenching effect was negligible. No radioactive liquid waste was generated by this method. In addition, you can reuse the measured sample.     

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%.     

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.     

Development of a practical decontamination procedure for uranium-contaminated concrete waste

Practical decontamination procedure and equipments have been developed to decontaminate uranium-contaminated concrete waste generated from a uranium conversion plant. The direct burning of mortar block coated with epoxy and then the mechanical removal of the burned surface reduces the amount of sludge by preventing the dissolution of whole cement paste in concrete block. And the removal of calcium from the concrete washing solution for the use of electrokinetic equipment decreases the volume of waste by a recycling of the solution. This improved decontamination procedure makes more than 70 % of concrete waste be self-disposed and remarkably reduces secondary radioactive waste.     

Waste minimization in the Los Alamos Medical Radioisotope Program

Since the mid-1970s the Los Alamos Medical Radioisotope Program has been irradiating target materials to produce and recover radioisotopes for applications in medicine, environmental science, biology, physics, materials research, and other disciplines where radiotracers find utility. By necessity, the chemical processing of targets and the isolation of radioisotopes generates radioactive waste materials. In recent years there have been federal mandates requiring us to discontinue the use of hazardous materials and to minimize radioactive waste volumes. As a result, substantial waste reduction measures have been introduced at the irradiation facility, in processing approaches, and even in the ways the product isotopes are supplied to users.     

Extraction fluorimetric determination of uranium by conventional spectrophotometry

The determination of uranium by a fluorimetric method using a conventional spectrophotometer has been elaborated. The quenching effect of the matrix was reduced by separation with liquid-liquid extraction and emulsion liquid membrane extraction methods using D2EHPA as a selective extraction reagent. The method was employed for uranium determination in radioactive waste solutions and proved to be very fast and easy to perform. It was found that it is possible to determinate as low as 0.2 ppm of uranium in a 10 ml sample.     

Limonene in Arizona liquid systems used in countercurrent chromatography. I Physicochemical properties

Limonene is a biorenewable cycloterpene solvent derived from orange peel waste. Its potential as a “green” solvent to replace heptane was recently evaluated. Countercurrent chromatography (CCC) is a preparative separation technique using biphasic liquid systems. One liquid phase is the mobile phase; the other liquid phase is the stationary phase held in place by centrifugal fields. A particular range of special proportions of the heptane/ethyl acetate/methanol/water system is called the Arizona (AZ) liquid system when the heptane/ethyl acetate ratio is exactly the same as the methanol/water ratio. A continuous polarity decrease is obtained between the most polar A composition (ethyl acetate/water or 0/1/0/1 v/v) and the least polar Z composition (heptane/methanol or 1/0/1/0 v/v), replacing heptane by limonene and methanol by ethanol produce biphasic liquid systems much more environmentallyfriendly than the original AZ compositions. The chemical compositions of the two liquid phases of 12 AZ limonene/ethyl acetate/ethanol/water proportions were fully determined by Karl-Fisher titration of water and by gas chromatography for the organic solvents. The results were compared with the compositions of the corresponding AZ mixtures containing heptane and methanol. Significant differences in ethyl acetate and ethanol distribution between phases of the two systems with identical volume proportions were established. The ratio of the upper phase over the lower phase volumes and the phase density difference are important in CCC, there are also significant differences between the classic and “green” AZ systems that are discussed.     

Extraction of uranium into a third phase of thorium nitrate—tributyl phosphate

Extraction of 0.05–0.25M uranyl nitrate into 30% tributyl phosphate (TBP) in dodecane from nitric acid solutions of thorium nitrate at equilibrium with its salt has been studied. Under investigated conditions a third (second organic) phase is formed. As the heavy organic phase extracts uranium, the calculated ratio of TBP to thorium and uranium sum decreases from 2.7 to less than 7. Electronic spectra show that in heavy organic phase approximately 80% of uranium is found as trinitrate complex, while in the light organic phase this complex is not detected. The measurements of dielectric constant () of the heavy phase reveal a frequency dependence of . The data obtained point to the existence of an ordered structure in the heavy organic phase.     

水/有机溶剂双相体系中色氨酸合成酶酶法合成L-色氨酸

在水/有机溶剂双相体系中,利用重组大肠杆菌DM206 [pET28a-trpBA/BL21 (DE3)]色氨酸合成酶催化L-丝氨酸和吲哚合成了L-色氨酸.考察了有机溶剂、有机相乙酸乙酯体积分数、反应温度、水相pH、底物L-丝氨酸与吲哚摩尔比、底物L-丝氨酸浓度、表面活性剂和酶添加量对色氨酸合成酶酶活的影响.结果表明,酶...     

Ion exchange in organic solvents

Studies on the partitioning of plutonium from 30% TBP by ion-exchange absorption on macroporous cation exchanger Amberlyst-15 have been described. Detailed loading experiments indicate that the resin absorbs plutonium in preference to uranium from loaded organic phase at low organic phase acidities (around 0.2M). Absorption behaviour of some fission products on the resin in 30% TBP is also reported. Possibility of using this procedure as an alternate method for plutonium partitioning from IAP stream of Purex process has been discussed.