A Bulk Resin Bead Procedure To Obtain Uranium And Plutonium From Radioactive Solutions For Mass Spectrometric Analysis

Abstract

A method is described for extracting representative uranium and plutonium samples from highly radioactive solutions for isotopic mass spectrometric analysis. Anion resin beads in the nitrate form are used to effect separation from fission products and other actinides. Conditions required to achieve separation are proper adjustment of the uranium and nitric acid concentrations. Once uranium and plutonium are adsorbed, each bead serves as a sample for mass spectrometric analysis, with plutonium and uranium being run sequentially from the same bead. Quantitative determination of the two elements is effected through the technique of isotopic dilution.     

The determination of plutonium by mass spectrometry using a [242]-plutonium tracer

An isotopic dilution method is described for the determination ot plutonium in samples of irradiated uranium using a ²⁴²Pu tracer. An aliquot of tracer is added to the sample and the mixture treated to ensure isotopic exchange; plutonium is then separated by an ion exchange procedure and an isotopic analysis made using an M.S.5. mass spectrometer. The precision (3 σ) for an aliquot containing 0.1 μg plutonium is 0.6%. A possible application of the method would be its use for control analyses of the feed solution in a chemical plant processing natural uranium fuel elements as, for example, the Windscale primary separation plant.     

Extraction of plutonium from sulphuric acid by TOA in toluene

The extraction of plutonium(VI) and plutonium(III) from sulphuric acid by TOA in toluene has been studied as a function of the acid and tri-octyl amine concentration. A comparison of the extraction properties of plutonium with those of uranium(VI) and uranium(IV) has been made. It was found that the extraction properties of plutonium(VI) are very similar to those of uranium(VI) and that TOA is a relatively poor extractant for plutonium(III). Uranium(IV) shows better extraction properties than plutonium(III). The results obtained are considered in the light of the stabilities of the complexes formed by these elements in the organic and aqueous phase. A method of separation of both elements by solvent extraction based on changing their oxidation states is suggested.     

萃取色层分离同位素稀释ICP-MS测定空气中费克 量钚

ICP-MS测定环境样品中超痕量^2^3^9Pu时,^2^3^8UH^+会对m/z239的测量带来干扰。测得UH^+的产生几率为4.6×10^-^5,通过三正辛胺色层分离后,对铀的去除率为10^4,可以有效地去除^2^3^8UH^+离子峰对^2^3^9Pu测定的干扰。钚的回收率为75%。同位素稀释法对^2^3^9Pu的检出限为4.5×10^-^1^5g/mL,方法的定量测定限为16×10^-^1^5g/mL。用所建立的方法测得我国某地区空气中^2^3^9Pu的浓度为4.8×10^-^1^7g/m^3。     

Solvent extraction studies on the recovery of plutonium(IV) from phosphoric acid solution using monooctylphenylphosphoric acid

Extraction of plutonium from analytical waste solutions containing phosphoric acid using a solution of monooctylphenylphosphoric acid (MOPPA) is described. Effect of reagent concentration, presence of uranium, plutonium loading and back extraction of plutonium from the organic phase are described. Using a solution of 0.05 F MOPPA in xylene, more than 90% plutonium could be extracted in one cycle.     

Preparation, certification and validation of a stable solid spike of uranium and plutonium coated with a cellulose derivative for the measurement of uranium and plutonium content in dissolved nuclear fuel by isotope dilution mass spectrometry.

A stable solid spike for the measurement of uranium and plutonium content in nitric acid solutions of spent nuclear fuel by isotope dilution mass spectrometry has been prepared at the European Commission Institute for Reference Materials and Measurements in Belgium. The spike contains about 50 mg of uranium with a 19.838% (235)U enrichment and 2 mg of plutonium with a 97.766% (239)Pu abundance in each individual ampoule. The dried materials were covered with a thin film of cellulose acetate butyrate as a protective organic stabilizer to resist shocks encountered during transportation and to eliminate flaking-off during long-term storage. It was found that the cellulose acetate butyrate has good characteristics, maintaining a thin film for a long time, but readily dissolving on heating with nitric acid solution. The solid spike containing cellulose acetate butyrate was certified as a reference material with certified quantities: (235)U and (239)Pu amounts and uranium and plutonium amount ratios, and was validated by analyzing spent fuel dissolver solutions of the Tokai reprocessing plant in Japan. This paper describes the preparation, certification and validation of the solid spike coated with a cellulose derivative.     

Extraction behavior of uranium,plutonium and some fission products with gamma-irradiated N,N′-dialkylamides

The extraction behavior of uranium(VI), plutonium(IV) and fission products like zirconium, ruthenium and europium from 3.5M nitric acid medium with gamma-irradiated dibutyl derivatives of hexanamide (DBHA), octanamide (DBOA) and decanamide (DBDA) in dodecane has been investigated as a function of absorbed dose up to 184 MRads. The results indicate that the K_d value for extraction of uranium(VI) decreases gradually, while K_d for extraction of plutonium(IV) decreases rapidly with dose up to 35 MRads, increasing thereafter with dose, indicating synergistic effects of radiolytic products at higher doses. Ruthenium and europium are not extracted in the entire dose range up to 184 MRads, while extraction of zirconium(IV) increases steadily up to 50 MRads and increases radiply thereafter, indicating synergistic effect of radiolytic products similar to that of plutonium(IV) beyond a dose of 50 MRads. The extractability of uranium(VI) and plutonium(IV) with 1M dibutyl decanamide (DBDA) in dodecane was studied for uranium loading up to 75 mg/ml and plutonium loading up to 3 mg/ml. The percent extraction was found to vary from 91 to 71 for uranium and 95 to 89 for plutonium, respectively. Quantitative stripping of uranium can be achieved with 0.01M nitric acid and plutonium with 0.5M nitric acid and 0.05M hydroxylamine soluton in two steps from an organic phase loaded with 53.2 mg/ml of uranium.     

Steric aspects of selectivity of organophosphorus extractants

This paper is devoted to the steric aspects of the efficiency and selectivity of extractants in the process of extraction of metal cations from acidic aqueous solution into a nonpolar phase. The extraction of uranium, plutonium, americium, and europium by tertiary carbamoylmethyl phosphine oxides and polydiphenylphosphinylmethyl benzenes have been studied. The separation factors of uranium and plutonium from americium are measured. Tertiary carbamoylmethyl phosphine oxides and diphenylphosphinylmethyl benzenes of the ortho type are characterized by unusually high separation factors. The diphenylphosphinylmethyl derivatives of benzene of the meta type do not possess a high selectivity in the process of separation of uranium and plutonium from americium, but are more efficient in extraction of americium.     

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.     

The chemical analysis of ternary alloys of plutonium with Molybdenum and Uranium

It is shown that the absorptiometric determination of molybdenum as thiocyanate may be used in the presence of plutonium. Molybdenum interferes with previously published methods for determining uranium and plutonium but conditions have been established for its complete removal by solvent extraction of the compound with a-benzoin oxime. The previous methods for uranium and plutonium are satisfactory when applied to the residual aqueous phase following this solvent extraction.     

Nuclear oxide fuel analysis

As a rule the analysis of nuclear oxide fuel includes the determination of uranium, plutonium, their isotopic composition, cation impurities, carbon, nitrogen, chlorine, fluorine, oxygen coefficient. In this paper we discuss different methods for the a analysis of unirradiated uranium and plutonium oxide fuelds used in the laboratories of the Analytical Chemistry Department: coulometry, emission analysis, chromatography, X-ray analysis. Much consideration is being given to the analysis of uranium and plutonium oxide samples and uranium-plutonium mixed fuels irradiated in the BOR-60 using mass-spectrometric (isotope dilution method) and radiometric techniques. The results of uranium and plutonium determination by these methods are compared. The main analytical characteristics of the methods are given.     

Extraction of uranium and plutonium from oxalate bearing solutions using phosphonic acid: Solvent extraction, extraction chromatography and infrared studies

Solvent extraction and extraction chromatography studies of uranium and plutonium from oxalate supernatant solutions were carried out using 2-ethyl hexyl-2-ethyl hexyl phosphonic acid (PC88A). Based on the distribution data, it was inferred that both the uranium and plutonium could be recovered satisfactorily from such a solution. These studies were found to be useful in optimising the appropriate concentration of PC88A, HNO₃, oxalic acid and temperature to recover more than 90% of plutonium from the large volumes of oxalate bearing waste solutions. Spectral characteristics of the extractant and its complexing behavior with U(VI) was also studied using IR & FTIR.     

Extractive photometric determination of plutonium(IV) with Aliquat-336 and xylenol orange

A rapid extractive photometric method using Aliquat-336 and xylenol organe for the determination of plutonium(IV) at μg levels has been developed. Quantitative extraction is obtained from ∼4M aqueous HNO₃ medium, affording estimation in the presence of several commonly occurring impurities, viz. iron, uranium, fission products and cladding materials. Effects of acidity, reagent concentration and diverse ions on the estimation have also been invetigated. Unlike the well-known absorptiometric method for determining plutonium(IV) employing Arsenazo III, the procedure presented here tolerates manyfold excesses of uranium(VI) as well as chromium(III), iron(III) and zirconium(IV), which are some of the major contaminants of plutonium during reprocessing.     

Determination of K-factors for isotope abundance measurements of uranium and plutonium by thermal ionisation mass spectrometry

K-factors (= certified isotope ratio/observed isotope ratio) are determined for the isotope abundance measurements of uranium and plutonium by thermal ionisation mass spectrometry. An mdf of 0.07% and 0.18% per mass unit differing by a factor of about 3, is obtained for uranium and plutonium, respectively, employing double rhenium filament assembly in the ion source and Faraday cup as the detector using the presently available isotopic reference materials of uranium and plutonium.     

Dioctyl butyramide and dioctyl isobutyramide as extractants for uranium(VI) and plutonium(IV)

Two isomeric monoamides, dioctyl butyramide (DOBA) and dioctyl isobutyramide (DOIBA) were synthesized for extracting uranium(VI) and plutonium(IV) from aqueous nitric acid medium into various diluents such asn-dodecane, tertiary butyl benzene and xylene. DOBA extracted uranium(VI) and plutonium(IV) efficiently whereas DOIBA extracted uranium(VI) with negligible extraction for plutonium(IV). Both these cations were extracted as their disolvates. The thermodynamic parameters involved in the extraction determined by the temperature variation method indicated the reactions in all cases to be enthalpy favoured and entropy disfavoured. Possibility of separating micrograms of plutonium(IV) from macroquantities of uranium(VI) using the mixture of these amides was explored.     

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.     

A quick method to determine uranium concentration by gamma spectroscopy: Its application for extraction of uranium from wet phosphoric acid

A rapid and accurate method to measure uranium was established using gamma spectroscopy with a high purity germanium detector. This can be done by measuring the gamma radiation intensity at 186 keV which corresponds to the concentration of ²³⁵U only since all the ²²⁶Ra which also shows a peak at this energy level, remains in the solid phosphogypsum. The method was proved to be accurate when compared with other methods. It has also the advantage of measuring uranium concentrations over a wide range suitable to the operation of the pilot plant for extraction of uranium from wet process phosphoric acid without the need to dilution or concentration. It can also measure the uranium in the different media such as acid, organic or alkaline media found in the pilot plant at different stages.     

Simultaneous Plutonium and Uranium Isotopic Analysis from a Single Resin Bead - A Simplified Chemical Technique for Assaying Spent Reactor Fuels

Abstract

The method uses basic anion resin to adsorb plutonium and uranium from 7–8 M HNO₃ solutions containing dissolved spent reactor fuels. After equilibrating the resin with the solution, a single bead is used to determine the isotopic composition of plutonium and uranium on sample sizes as small as 10^?9 to 10^?10 g of each element per bead. Isotopic measurements are essentially free of isobaric interferences and fission product contamination in the mass spectrometer is eliminated. A very small aliquot of dissolver solution containing 10^?6 g of U and 10^?8 g of Pu is sufficient sample for chemically preparing several resin beads. A single prepared bead is loaded onto a rhenium filament and analyzed in a two-stage mass spectrometer using pulse counting for ion detection to obtain the high sensitivity required. Total quantity of the elements, in addition to isotopic abundances, can be determined by isotope dilution. Other areas where the method may be useful are: in plutonium production, isotope separations, and for trace detection of contamination on reactor parts.     

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.     

Thermal hazard analysis of mixed system of hydrazine and nitric acid

In order to obtain a better understanding of the thermal hazard of hydrazine and nitric acid mixtures which are used in plutonium and uranium recovery by extraction (PUREX) process during recycling of spent fuel in nuclear plant, DSC was carried out for the mixed system made by the small scale reaction calorimeter super-CRC in various conditions. It was found that the concentration of the nitric acid has a strong influence on the thermal behavior of the mixed system, and hydrazine nitrate which is formed in the reaction in the mixture was determined by DSC with FTIR analysis.