Uranium/plutonium and uranium/neptunium separation by the Purex process using hydroxyurea

Hydroxyurea dissolved in nitric acid can strip plutonium and neptunium from tri-butyl phosphate efficiently and has little influence on the uranium distribution between the two phases. Simulating the 1B contactor of the Purex process by hydroxyurea with nitric acid solution as a stripping agent, the separation factors of uranium/plutonium and uranium/neptunium can reach values as high as 4.7·10⁴ and 260, respectively. This indicates that hydroxyurea is a promising salt free agent for uranium/plutonium and uranium/neptunium separations.     

Uranium/plutonium and uranium/neptunium separation by the Purex process using hydroxyurea

Hydroxyurea dissolved in nitric acid can strip plutonium and neptunium from tri-butyl phosphate efficiently and has little influence on the uranium distribution between the two phases. Simulating the 1B contactor of the Purex process by hydroxyurea with nitric acid solution as a stripping agent, the separation factors of uranium/plutonium and uranium/neptunium can reach values as high as 4.7·10⁴ and 260, respectively. This indicates that hydroxyurea is a promising salt free agent for uranium/plutonium and uranium/neptunium separations.     

Individual determination of uranium and plutonium in a single aliquot

Studies on the individual potentiometric determination of uranium and plutonium in a single aliquot have been initiated recently in our laboratory. It was required to adapt the reported procedures (for the precise determination of uranium and plutonium individually when present together in a sample) at various stages to make them suitable for the successive application of the procedures to the same aliquot. Two alternative schemes are proposed in the present work. In the first, plutonium is determined by HClO₄ oxidation followed by the determination of total uranium and plutonium by Zn(Hg) reduction. In the second, plutonium is determined by AgO oxidation following the determination of total uranium and plutonium by Zn(Hg) reduction. Amount of uranium is computed in both cases from the difference of two determinations. Precision for the assay of plutonium and uranium was found to be ±0.25% and ±0.35%, respectively, at milligram levels.     

Extraction of uranium(VI) and plutonium(IV) with some aliphatic amides

Extraction of uranium(VI) and plutonium(IV) has been studied with N,N-dibutyl derivatives of hexanamide (DBHA), octanamide (DBOA) and decanamide (DBDA) at various fixed temperatures of 20, 30, 40 and (50±0.1)°C. The equilibrium constants for the uptake of nitric acid (K_h, a measure of their relative basicities) by these amides were evaluated by the usual method. The equilibrium constants for the extraction of uranium as well as plutonium with all the three amides follow their order of basicity (K_h) viz. DBHA (0.09)<DBOA (0.10)<DBDA (0.13) with log K values of 1.31, 1.43 and 1.73 for uranium and 3.55, 3.65 and 4.17 for plutonium, respectively. It has been observed that whereas uranium(VI) is extracted as a disolvate (similar to TBP and sulfoxides), plutonium(IV) has been found to be extracted as a trisolvate. The thermodynamic parameters evaluated by the usual temperature coefficient method indicate that the extraction reactions of uranium as well as plutonium are stabilized by negative enthalpy change only.     

Simultaneous determination of uranium and plutonium at trace levels in process streams using Arsenazo III by derivative spectrophotometry

A derivative spectrophotometric method has been developed for the simultaneous determination of uranium and plutonium at trace levels in various process streams in 3M HNO₃ medium using Arsenazo III. The method was developed with the objective of measuring both uranium and plutonium in the same aliquot in fairly high burn-up fuels. The first derivative absorbances of the uranium and plutonium Arsenazo III complexes at 632 nm and 606.5 nm, respectively, were used for their quantification. Mixed aliquots of uranium (20–28 μg/ml) and plutonium (0.5–1.5 μg/ml) with U/Pu ratio varying from 25 to 40 were analysed using this technique. A relative error of about 5% was obtained for uranium and plutonium. The method is simple, fast and does not require separation of uranium and plutonium. The effect of presence of many fission products, corrosion products and complexing anions on determination of uranium and plutonium was also studied.     

Determination of uranium in plutonium by differential linear sweep oscillographic polarography

The polarographic behavior of uranium in hydroxylamine hydrochloride was investigated by differential oscillographic polarography. A procedure is presented for the determination of uranium in plutonium for concentrations of uranium greater than 10 p.p.m. Analyses of solutions containing 22 common impurities found in plutonium metal revealed that antimony, copper, and titanium cause significant interference. A reversible peak corresponding to a one-electron reduction was obtained with a peak potential of -0.167 V vs. Hg pool electrode. The diffusion coefficient is 0.51·10^-5 cm²/sec and the diffusion current constant is 1.59 with an average relative standard deviation of 2.28%. The peak current of uranium can be affected by hydrochloric, nitric, perchloric, and sulfuric acids, depending on the acid concentration.     

Determination of plutonium in the presence of uranium by alpha-spectrometry

The plutonium determination by alpha-particle spectrometry with semiconductor detectors in the presence of uranium has been described. It has been found that plutonium as well as uranium can be electrodeposited quantitatively on nickel or stainless steel discs from solutions in isopropanol. The time of deposition does not exceed 35–40 min. The determination of plutonium is possible within the uranium to plutonium weight ratio of 4000 with the accuracy better than 2%.     

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 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.     

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.     

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.     

A ratio derivative spectrophotometric method for the simultaneous determination of uranium and plutonium

A ratio derivative spectrophotometric method has been developed for the simultaneous determination of uranium and plutonium at mg levels in 1M HNO₃ medium. In this method the overlapping spectra of uranium and plutonium are well resolved by making use of the first derivative of the ratios of their direct absorption spectra. The derivative ratio absorbances of uranium and plutonium are measured at 411.2 and 473.8 nm, respectively for their quantification. The method is simple, fast and does not require separation of uranium and plutonium. Another salient feature of the method is that it does not lead to generation of analytical waste thereby minimising the efforts required for the recovery of plutonium. Uranium in the conc. range of 10–25 mg/g and plutonium in the conc. range of 0.5 to 2 mg/g (U/Pu ratio varying from about 10 to 25) were determined in the same aliquot with a precision and accuracy of about 0.5% and 1%, respectively.     

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.     

Synthesis of dihydroxyurea and its application to the U/Pu split in the PUREX process

Dihydroxyurea (DHU) was synthesized using tri-associated solid phosgene [bis(trichloromethyl) carbonate] dissolved in dioxane and hydroxylamine hydrochloride dissolved in potassium acetate solution. The reduction of Pu(IV) by DHU was investigated using UV-Vis spectrophotometry. The reduction back-extraction behavior of Pu(IV) in 30% tri-butyl phosphate/kerosene was firstly investigated under conditions of various temperature, various DHU and HNO₃ concentrations and various phase contact times. The results showed that Pu(IV) in the organic phase can be stripped rapidly to the aqueous phase by DHU. Simulating the 1B contactor of the PUREX process using a 0.1 M DHU in 0.36M nitric acid solution as the stripping agent, the separation factors of uranium/plutonium can reach 2.1·10⁴. This indicates that DHU is a promising salt free agent for uranium/plutonium separation.     

Simultaneous determinations of alpha-emitting isotopes of uranium and plutonium in bone

A radiochemical procedure has been developed for the simultaneous determinations of alpha-emitting isotopes of uranium and plutonium in bovine bone. The radiochemical recoveries ranged from 61 to 93% with a mean of 80±10% for uranium and 57 to 88% with a mean of 75±11% for plutonium. The method was successfully used for human bone.     

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.     

Coulometric determination of uranium and iron or plutonium in a single aliquot using platinum working electrode

A controlled potential coulometric method developed earlier for the determination of uranium in the presence of iron or plutonium using platinum working electrode was extended for individual determination of uranium and iron or plutonium in single aliquot. After uranium determination, Fe(III) or Pu(IV) in the aliquot is reduced electrolytically to Fe(II) or Pu(III) and subsequently determined by electrolytic oxidation to Fe(III) or Pu(IV), respectively. Analysis of synthetic solutions indicated that the values for uranium, iron and plutonium obtained by this method are reproducible within±0.2% and are in good agreement with values obtained using conventional redox methods 1, 2.     

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.     

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.     

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.