Complexation of uranium(VI) with acetohydroxamic acid

The advanced separation extraction process based on tri-n-butyl phosphate organic phase called UREX is being developed to separate uranium from fission products and other actinides, and the acetohydroxamic acid (AHA) is employed to reduce and complex plutonium and neptunium in order to decrease their distribution to the TBP-organic phase. In this study, the extraction of uranium was performed from various aqueous matrices with different concentrations of HNO₃, LiNO₃, and AHA. Extraction of uranium increases with increasing both initial HNO₃ and total nitrate concentration. UV-VIS spectrophotometry confirmed that AHA is involved in the complex of uranium with TBP.     

Polarographic behavior of uranium(VI) in malonic acid media determination of uranium in plutonium

The polarographic behavior of the uranium-malonate complex was investigated over the pH range 1.1–6.5. A reversible, one-electron wave was obtained. Below pH 4.9, the rate of disproportionation is nearly instantaneous and gives rise to a pseudo uranium(VI)-uranium(IV) reduction. Above pH 4.8 the concentration of uranium(V) is stable with respect to disproportionation. The half-wave potential is pH-dependent below pH 4.9, but it is independent of the malonate concentration above O.1 M. The diffusion current constant is 2.78 for the conditions described. A procedure for the determination of uranium in plutonium was developed for uranium concentrations greater than 225 p.p.m. Of 21 common impurities found in plutonium metal, only Cu, Fe, Pb, Sb and Ti cause significant interference ; titanium can be removed by ion exchange, and the other interferences by mercury cathode electrolysis.     

Extraction of uranium(VI) and plutonium(IV) with dihexylbutyramide and dihexylisobutyramide from nitric acid medium

The extraction of uranium(VI) and plutonium(IV) was carried out with two isomeric monoamides, dihexylbutyramide (DHBA) and dihexylisobutyramide (DHIBA) from nitric acid medium, usingn-dodecane as diluent. The possibility of separation of the two metal ions from each other without valency adjustment was attempted. U(VI) was extracted as its disolvate, while Pu(IV) was extracted as its trisolvate. From the variation of distribution ratio with temperature, it was shown that the extraction reaction was enthalpy controlled in all the cases.     

Reduction of pertechnetate by acetohydroxamic acid: formation of [TcII(NO)(AHA)2(H2O)]+ and implications for the UREX process

Reductive nitrosylation and complexation of ammonium pertechnetate by acetohydroxamic acid has been achieved in aqueous nitric and perchloric acid solutions. The kinetics of the reaction depend on the relative concentrations of the reaction components and are accelerated at higher temperatures. The reaction does not occur unless conditions are acidic. Analysis of the X-ray absorption fine structure spectroscopic data is consistent with a pseudo-octahedral geometry and the linear Tc-N-O bond typical of technetium nitrosyl compounds, and electron spin resonance spectroscopy is consistent with a d (5) Tc(II) nitrosyl complex. The nitrosyl source is generally AHA, but it may be augmented by some products of the reaction with nitric acid. The resulting low-valency trans-aquonitrosyl(diacetohydroxamic)-technetium(II) complex ([Tc (II)(NO)(AHA) 2H 2O] (+), 1) is highly soluble in water, extremely hydrophilic, and is not extracted by tri- n-butylphosphate in a dodecane diluent. Its extraction properties are not pH-dependent: potentiometric-spectrophotometric titration studies indicate a single species from pH 4 down to -0.6 (calculated). This molecule is resistant to oxidation by H 2O 2, even at high pH, and can undergo substitution to form other technetium nitrosyl complexes. The potential formation of 1 during reprocessing may strongly impact the fate of technetium in the nuclear fuel cycle.     

Extraction of uranium(VI) and plutonium(IV) with some high molecular weight aliphatic monoamides from nitric acid medium

The extraction behavior of U(VI) and Pu(IV) with dioctyloctanamide (DOOA), dioctylethylhexanamide (DOEHA) and diisobutylethylhexanamide (DIBEHA) was investigated from nitric acid medium. With DOOA, U(VI) extraction is higher than that for Pu(IV) upto 5M HNO₃ and the trend is reversed at higher acid concentrations. Extraction yield of U(VI) is higher than that for Pu(IV) in the case of DOEHA and DIBEHA. DIBEHA extraction of Pu(IV) is found to be very small. The lower value of the distribution ratio for Pu(IV) with branched amides was attributed to steric reasons. The possibility of using these amides for separation of U(VI) and Pu(IV) without valency adjustment was explored. Both U(VI) and Pu(IV) are extracted as their disolvates by DOOA and DOEHA.     

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.     

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.     

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.     

Determination of uranium and plutonium in plutonium based fuels by sequential amperometric titration

A method is described for the sequential determination of uranium and plutonium in plutonium bearing fuel materials. Uranium and plutonium are reduced to U(IV) and Pu(III) with titanous chloride and then titrated with dichromate to two end points which are detected amperometrically using two polarized platinum electrodes. Uranium-plutonium solutions of known concentrations containing plutonium in the proportions of 4, 30, 50, and 70% were analyzed with precisions better than 0.3%, maintaining the amounts of plutonium per aliquot in the range of 2–10 mg. No significant bias could be detected. Several samples of (U, Pu)O₂ and (U, Pu)C were analyzed by this procedure. The effects of iron, fluoride, oxalic acid and mellitic acid on the method were also studied.     

Plutonium process control using an on-line gamma monitor for uranium,plutonium, and Americium

An On-Line Gamma Monitor profiles the concentration of uranium, plutonium, and americium in waste and product streams of the anion exchange process used to purify plutonium at Los Alamos. The Monitor employs passive gamma Spectrometry to measure the 59.5-KeV and 129-KeV gamma rays of²⁴¹Am and²³⁹Pu, respectively. Because the uranium impurity in typical process streams has no gamma ray suitable for passive measurement, a novel radiotracer technique is used. Uranium-237, always present in plutonium processed at Los Alamos as a minor alpha-decay daughter of²⁴¹Pu, has a 6.8-day half-life and 208-KeV gamma energy, which make it an ideal radiotracer for macro amounts of uranium in the process. The On-Line Gamma Monitor is used routinely to provide Los Alamos operators with continuous, real-time process control information.     

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.     

On atomization of uranium in plutonium matrix using ETA-AAS

Atomization processes for uranium in aqueous media and in the presence of a plutonium matrix have been studied and a chemical mechanism is proposed. These studies have been utilized for the determination of uranium in plutonium by Electrothermal Atomization- Atomic Absorption Spectrometry (ETA-AAS) within the constraints of its stable carbide forming tendency and complexity caused by formation of plutonium suboxide at high temperatures. In spite of these limitations the analytical range obtained for determination of uranium is 2.5–100 ng with a sample aliquot of 5 μL containing 5 mg/mL plutonium concentration. The precision of the method is evaluated as 9% RSD. Received: 9 September 1997 / Revised: 29 December 1997 / Accepted: 31 December 1997     

Microwave-assisted efficient one-step synthesis of amides from ketones and benzoxazoles from (2-hydroxyaryl) ketones with acetohydroxamic acid using sulfuric acid as the catalyst

Efficient one-step method for the synthesis of amides directly from ketones and benzoxazoles from (2-hydroxyaryl) ketones by the reaction of acetohydroxamic acid using sulfuric acid as catalyst was described.     

Kinetics and mechanism of stripping of Np(IV) by acetohydroxamic acid using a Lewis cell

Kinetic studies of stripping of Np(IV) from 30% Tri-Butyl-Phosphate/Odourless Kerosene (TBP/OK) into a nitric acid solution containing acetohydroxamic acid (CH₃CONHOH) have been investigated using a Lewis cell. The different parameters affecting the back-extraction rate of Np(IV) such as Np, TBP, nitric acid, nitrate, acetohydroxamic acid(AHA) concentration in addition to temperature, stirring speed and special interfacial area were separately studied and a rate equation was deduced. Results have been compared among themselves and other published works on similar systems. Mechanisms of stripping processes have been proposed.     

Determination of uranium in presence of plutonium using differential spectrophotometry

The use of beams of heavy ions such as carbon, neon and argon for radiation therapy has the advantage that they have a very sharp Bragg maximum. When the Bragg peak coincide with the tumour location, it is possible to deposit the bulk of the energy of the ion in the region occupied by the malignancy. However, the concentration of ions and free radicals will be very high in the Bragg peak region which has not received the attention it deserves. So mutual recombination of these species will be very high. It is therefore necessary to assess the extent of these radical-radical recombinations at very high LET values. Spur diffusion model calculations have been made for high energy argon ions using water as a medium. For comparison, calculations have been done for proton tracks. It has been shown that in the Bragg peak region of argon ions even very high concentrations of scavengers have very little effect on radical-radical interactions. The implication is that when LET values are very high, practically all the radicals undegro recombination with each other. In order to explain the observed lethality of high LET radiation, it is suggested that the hydrogen peroxide formed also contribute to the killing of cells. In addition, the decomposition of H₂O₂ will contribute oxygen. This may be one of the reasons why high LET radiation shows strong lethality to hypoxic cells.     

Purification of americium from bulk quantities of plutonium loading effluent using trilaurylamine

Four litres of Am solution containing 81.5 mg/l Am and 0.6 mg/l Pu could be purified by a two-step procedure involving solvent extraction and extraction chromatography with trilaurylamine. The final product contained a maximum concentration of 9 g/l Pu.