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.     

Extraction of plutonium(IV), uranium(VI) and some fission products by di-n-hexyl sulphoxide

The extraction of nitric acid, plutonium, uranium and fission products such as zirconium, ruthenium and europium has been investigated using di-n-hexyl sulphoxide in Solvesso-100. Results indicate that Pu(IV), U(VI), Zr(IV) and Ru NO(III) are extracted as disolvates, whereas Eu(III) is extracted as the trisolvate. The absorption spectra of the plutonium(IV) and uranium(VI) complexes extracted are similar to those of the species extracted by TBP which indicate the similarity of the species involved. Preliminary studies show that irradiated di-n-hexyl sulphoxide extracts zirconium to a smaller extent than irradiated TBP suggesting the use of long chain aliphatic sulphoxides as promising extractants for the recovery of plutonium in high radiation fields.     

Purification of uranium product from plutonium contamination using acetohydroxamic acid (AHA) based process

Acetohydroxamic acid (AHA) based uranium product purification process to remove plutonium was optimized. For this process, equilibrium data was generated to optimize AHA concentration and acidity of stripping agent/scrubbing agent. Two options namely (i) Pu complexation in aqueous phase followed by extraction and scrubbing ii) extraction followed by scrubbing with AHA were studied. Results of these studies indicate that U product obtained in AHA purification is near to the table top specification and also quantitative Pu recovery from the AHA strip product is possible by oxalate precipitation.     

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.     

Controlled potential coulometry: the application of a secondary reaction to the determination of plutonium and uranium at a solid electrode

A method is described for the simultaneous determination of plutonium and uranium in mixed oxides by controlled potential coulometry at a gold working electrode in two stages: first a coulometric oxidation, at 0.73 V vs. a silver/silver chloride electrode, of Pu(III) and U(IV) to Pu(IV) and U(VI) by a combination of a direct electrode reaction and a secondary chemical reaction proceeding concurrently, and secondly, a coulometric reduction at 0.33 V of Pu(IV) to Pu(III), leaving uranium as U(VI). The determination is carried out in a mixture of sulphuric and nitric acids, and Ti(III) is used to reduce plutonium and uranium to Pu(III) and U(IV) before electrolysis. The precision (3sigma) of Pu:U ratio results obtained from mixtures containing about 30% and 2% plutonium was 0.5% and 1-5% respectively. The effect of experimental variables on the time taken to complete the coulometric determination is discussed.     

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.     

Comparing Branched versus Straight-chained Monoamide Extractants for Actinide Recovery

Separations of used nuclear fuel at the engineered scale have generally been completed using the Plutonium Uranium Redox Extraction (PUREX) process. The PUREX process uses tributyl phosphate (TBP) as an extractant to recover uranium and plutonium. While the TBP extractant has proven effective at recovering U and Pu at the engineered scale, TBP is potentially vulnerable to third phase formation and TBP degradation products (monobutyl and dibutyl phosphoric acids) which can complicate recovery of extracted metals from the organic phase. An alternative class of extractants, monoamides, has been considered for applications in thorium and uranium fuel cycles. When compared to TBP, monoamides tend to have higher separation factors for U or Pu from fission products, structural materials, and Th. This review summarizes the literature that explores actinide separations using monoamides by assessing the physiochemical properties between a broader library of branched and straight-chain monoamides than considered in previous reviews. An emphasis is placed on fine-tuning the selectivity of branched monoamides.     

Determination of plutonium and uranium in mixed oxide fuels by sequential redox titration

The use of a sequential determination of uranium and plutonium in a single sample solution results in a saving in analysis time and apparatus requirements. The method starts with U(IV) and Pu(in) in a mixture of sulphuric and nitric adds. Titration with dichromate, using amperometry at a pair of polarizable electrodes, produces two well-defined end-points corresponding to the sequential oxidation of U(IV) to U(VI) and Pu(III) to Pu(IV). The quantitative oxidation of U(IV) to U(VI) is achieved via the action of Pu(IV) as intermediate, and is dependent upon establishing conditions which favour rapid reaction between U(IV) and Pu(IV). The method is precise and accurate. With Pu-U mixtures containing between 15 and 30% plutonium the precision (3sigma) of the Pu: U ratio results is +/-0.6% on samples containing 100-120 mg of plutonium plus uranium. Iron and vanadium interfere quantitatively with plutonium, copper interferes non-quantitatively with uranium, and gross amounts of molybdenum mask the uranium end-point.     

Alpha spectroscopic analysis of actinides (Th,U and Pu) after separation from aqueous solutions by cation-exchange and liquid extraction

The radioactivity concentration of ²³⁶Pu, ²³²U and ²²⁸Th in aqueous samples has been determined by means of alpha spectroscopy after chemical separation and pre-concentration of the radionuclides by cation exchange and liquid–liquid extraction using the Chelex-100 resin and 30% TBP/dodecan, respectively. Method calibration using a ²³⁶Pu standard solution containing the daughter radionuclides results in a detector efficiency of 18% and in a chemical recovery for cation-exchange which is (30 ± 7)%, (90 ± 5)% and (20 ± 5)% for plutonium, uranium and thorium, respectively. The chemical recovery for liquid–liquid extraction is found to be (60 ± 7)%, (50 ± 5)% and (70 ± 5)%, for plutonium, uranium and thorium, respectively. The differences in the efficiencies can be ascribed to the oxidation states, the different actinides present in solution. Taking into account that the electrodeposition of the radionuclides under study is quantitative, the total method efficiency is calculated to be (18 ± 15)%, (46 ± 7)% and (15 ± 5)%, for plutonium, uranium and thorium, respectively, at the mBq concentration range. The detection limit of the alpha spectrometric system has been found to be 0.2 mBq/L, suggesting that the method could be successfully applied for the radiometric analysis of the studied radionuclides and particularly uranium in aqueous samples.     

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.     

Effect of mixed solvent media on the sorption and separation of uranium and plutonium on macroporous resins

Ion-exchange studies on uranium and plutonium using macroporous (MP) anion-exchange resins from an aqueous-organic solvent mixed media were carried out to develop a separation method. Out of the several water miscible organic solvents tried methanol and acetone were found to be best suited. Distribution data were obtained for U(VI) and Pu(IV) using three macroporous resins under various parameters. Based on these data, separation factors for Pu/U were calculated. Column experiments using Tulsion A-27(MP) were also carried out. The method has the advantage of loading plutonium from as low as 1M nitric acid in the presence of methanol or acetone and could be used satisfactorily for its recovery from solutions containing plutonium and uranium.     

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.     

The application of <Emphasis Type="Italic">N</Emphasis>,<Emphasis Type="Italic">N</Emphasis>-dimethylhydroxylamine as reductant for the separation of plutonium from uranium

Both single stage and multi-stages experiments on stripping plutonium with N,N-dimethylhydroxylamine (DMHAN) as reductant with methylhydrozine (MMH) as supporting reductant were carried out. The effect of contact time, temperature, acidity, concentration of DMHAN on back-extraction rate of plutonium was investigated in the single stage experiment. The results demonstrated that the reaction of stripping Pu(IV) in the organic phase (30% TBP–kerosene) 1BF solutions by DMHAN exhibits excellent stripping efficiency. Under the given conditions, the back-extraction rate of plutonium reaches 90% within 2 min. Higher temperature, lower acidity and the increased concentration of DMHAN benifit the stripping reaction. The concentration profile of HNO₃, uranium and plutonium were determined in a multi-stages mixer-settler after the steady state of the back-extraction, and the multi-stages results show that the plutonium can be separated effectively from uranium. The recovery of plutonium and uranium reach 99.995% or over 99.99% respectively. The separation factor of U from Pu (SF_Pu/U) is about 2 × 10⁴.     

Redox determination of uranium in presence of plutonium in low phosphoric acid medium and the recovery of plutonium

Quantitative determination of uranium in (U, Pu)O₂ fuels is usually done by the DAVIES-GRAY method. High concentrations of phosphoric acid in the analytical waste generated by this method make the revocery of plutonium rather complex. Studies on the recovery of plutonium from nitric acid medium containing different concentrations of H₃PO₄ by conventional anion-exchange procedure reveal that more than 90% of the plutonium can be easily recovered when the phosphoric acid concentration is less than 0.5 M in the solution. A method was developed for the determination of uranium in the presence of plutonium, which involves the reduction of U(VI) to U(IV) by Fe(II) in a medium of 3.5M H₃PO₄ +4.5M H₂SO₄ instead of 10–11M H₃PO₄ so as to have the H₃PO₄ concentration 0.6M in the waste. A number of determinations of uranium in UO₂(NO₃)₂ working standard solutions and (U, Pu) synthetic solutions with uranium at the 3–7 mg level were carried out by this method. The precision obtained was better than ±0.2% and the accuracy was also within the precision limits. The resulting analytical waste generated was directly subjected to anion exchange separation for the recovery of plutonium which was found to be more than 90%.     

Separation of uranium from different uranium oxide matrices employing supercritical carbon dioxide extraction

Uranium from different uranium oxide matrices was extracted with tri-n-butyl phosphate–nitric acid (TBP–HNO₃) adduct using supercritical carbon dioxide (SC CO₂). While 30 min dissolution time at 323 K was sufficient for U₃O₈ and UO₂ powder, UO₂ granule (at 333 K) and crushed green pellet (at 353 K) required 40 min. Crushed sintered pellet required 60 min at 353 K for complete dissolution. Influence of various experimental parameters such as temperature, pressure, volume of TBP–HNO₃ adduct, acidity of nitric acid used for preparing TBP–HNO₃ adduct and extraction time on uranium extraction efficiency was also investigated. For UO₂ powder, temperature of 323 K, pressure of 15.2 MPa, 1 mL TBP–HNO₃ adduct, 10 M nitric acid and 30 min extraction time was found to be optimum. ~70% uranium extraction efficiency was obtained on extraction with SC CO₂ alone which increased to 90% with the addition of 2.5% TBP in SC CO₂ stream. Extraction efficiency was found to vary linearly with TBP percentage and nearly complete uranium extraction (~99%) was observed with 20% TBP. Nearly complete extraction was also achieved with addition of 2.5% thenoyltrifluoroacetylacetone (TTA) in methanol. The optimized procedure was extended to remove uranium from simulated tissue paper waste matrix smeared with uranium oxide solids.     

Kinetics and mechanism of stripping of Pu(IV) by <Emphasis Type="Italic">N</Emphasis>,<Emphasis Type="Italic">N</Emphasis>-dimethylhydroxylamine using a Lewis cell

The kinetics of back-extraction of Pu(IV) from 30% Tri-Butyl-Phosphate/Odourless Kerosene (TBP/OK) into a nitric acid solution containing N,N-dimethylhydroxylamine (DMHAN) have been investigated using a Lewis cell. The different parameters affecting the back-extraction rate of Pu(IV) such as plutonium, nitric acid, DMHAN concentration in addition to temperature, stirring speed were separately studied and a rate equation was deduced. The activation energy of this process is 28.4 kJ/mol.     

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.     

Alpha radiolytic degradation of <Emphasis Type="Italic">N,N</Emphasis>-dihexyl octanamide

The primary purpose of this study was to understand the alpha radiolytic degradation behavior of N,N-dihexyl octanamide (DHOA) vis a vis tributyl phosphate (TBP) solutions in n-dodecane under plutonium loading conditions. These studies were carried out as a function of dose on different Pu loaded samples (containing 0.002-10 g/L Pu) from 4 M HNO₃ medium. These Pu loaded solutions were evaluated for stripping behavior by contacting with 0.5 M NH₂OH at 0.5 M HNO₃ solutions. Organic phase analysis was carried out by gas chromatography (GC) and by visible spectrophotometry. These studies clearly indicated that Pu stripping becomes difficult with increased dose in the case of TBP system. On the other hand, no such problem was observed in DHOA system during stripping of plutonium, thereby indicating that DHOA is a promising candidate for the reprocessing of high burn up Pu rich spent fuels.     

Di(2-ethyl hexyl) butyramide and di(2-ethyl hexyl)isobutyramide as extractants for uranium(VI) and plutonium(IV)

The extraction of uranium(VI) and plutonium(IV) from nitric acid into n-dodecane was studied using two isomeric branched alkyl amides, di(2-ethyl hexyl) butyramide (DEHBA) and di(2-ethyl hexyl) isobutyramide (DEHIBA). The extraction ratios of Pu(IV) at relatively high acidities were higher than the corresponding values for U(VI) in the case of DEHBA. However, with DEHIBA the values for Pu(IV) were negligibly small. Pu(IV) was found to be extracted as trisolvate by DEHBA and as disolvate by DEHIBA. U(VI) was extracted by both the amides. From the study of the extraction reactions at different temperatures, it was shown that all the reactions in the present investigation were enthalpy favoured and entropy disfavoured. Separation of Pu(IV) from bulk of U(VI) was feasible. However, the purity of the separated plutonium was not satisfactory in batch extraction studies.     

Reduction kinetics of Pu(IV) and Np(VI) by N,N-dimethylhydrazine,and its potential application in nuclear fuel reprocessing

Reduction kinetics of Pu(IV) by N,N-dimethylhydrazine (NNDMH) were studied by spectrophotometry, and the reduction rate equation in 3M (mol/dm³) nitric acid was obtained. The reduction properties of NNDMH for U(VI), Np(VI), and Pu(IV) was studied in the mixture solution of trin-butylphosphate diluted to 30 vol.% by n-dodecane (30% TBP) and 3M nitric acid. It was confirmed that NNDMH selectively reduce Np(VI) to Np(V) without affecting the valences of U(VI) and Pu(IV) in a few minutes. Numerical simulation indicated that 99.9% of Np was separated from U and Pu applying NNDMH for a mixer-settler.