Studies on purification of plutonium from silver ions

Precipitation and solvent extraction methods have been investigated for the purification of plutonium from silver from the solution generated during oxidative dissolution of plutonium oxide using Ag(II) ions. Initial experiments have been carried out using thorium as representative of plutonium. Selecting the optimum conditions, the experiments were repeated with plutonium. The results revealed that Pu can be purified from silver ions either by precipitating silver as silver chloride or silver metal followed by Pu(IV) oxalate precipitation or by selective extraction of Pu(IV) into 20% Aliquat-336 or 30% TBP.     

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.     

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.     

Recovery of plutonium from carbonate wash solutions

A novel method has been developed for recovery of plutonium and uranium from carbonate wash solutions generated during solvent wash process involved in the reprocessing of high burn up FBTR fuel. The proposed method involves a selective coprecipitation of Pu and U by adding ammonium hydroxide to the pre acidified carbonate wash solution. Substantial removal of DBP by successive steps of coprecipitation, completely eliminates the possibility of undesired solid formation which is mainly due to the presence of high content of DBP. By adopting this method, an excellent decontamination factor for DBP has been achieved without any crud/solid formation. Phosphate content in the final oxide product meets the product specifications. Flowsheet condition necessary for the recovery process for plutonium from the aqueous carbonate solution is formulated and adopted in the CORAL facility.     

Studies on sorption of plutonium on inorganic ion exchangers from sodium carbonate medium

Sorption of Pu(IV) from sodium carbonate medium has been investigated by using three inorganic ion exchangers, viz. alumina, silica gel and hydrous titanium oxide (HTO). Distribution ratios (D) of Pu(IV) for its sorption on these ion exchangers have been determined. The values are 700, 10³ and 10⁴ for alumina, silica gel and hydrous titanium oxide, respectively, from 0.1M sodium carbonate medium. The high distribution ratios indicate their suitability for the removal of Pu(IV) from sodium carbonate waste streams. Pu(IV) breakthrough capacities have been determined with 5 ml bed at a flow rate of 30 ml per hour. The 10% Pu(IV) breakthough capacities for alumina and silica gel are 3 g l^–1 and 14 g l^–1, respectively. The capacity of HTO is 60 g of Pu(IV) per liter of exchanger at 4% Pu(IV) breakthrough.     

Analysis of plutonium in biological and environmental materials

A procedure for separation of plutonium from some biological and environmental materials has been tested in model and real conditions. The procedure involves a commonly used way of conversion of plutonium to oxidation state (IV) in nitric acid medium and sorption of Pu(IV) on a strongly basic anion exchanger from hydrochloric acid medium thus eliminating interference of²²⁸Th with the²³⁸Pu analysis.     

Spectrophotometric determination of plutonium III, IV, and VI concentrations in nitric acid solution

The determination of the concentration of various valency states of plutonium is desirable in various stages of the Plutonium/Uranium Recovery by EXtraction (PUREX) process for the effective separation and purification of plutonium. A method is optimized for the quantitative spectrophotometric determination of Pu(III), Pu(IV) and Pu(VI) existing separately or in mixed oxidation states in 1.5?M nitric acid medium. Molar absorption co-efficient (??) for the major absorption peaks with baseline correction are evaluated. With these ?? data a method is proposed for determining the molar concentration of each oxidation state.     

Sorption of Pu(IV) on Aliquat-336 impregnated silica-gel

A new ion exchange material prepared by impregnating Aliquat-336 on silica-gel has been investigated for the recovery of plutonium from nitric-oxalic acid solutions. The distribution ratio of Pu(IV) was studied at various concentrations of nitric and oxalic acids. The presence of Al(III) and Fe(III) in the solution, enhances the uptake of Pu(IV). Pu(IV) breakthrough capacities (btc) have been determined using 2.5 ml bed of the ion exchange material column in the absence and the presence of Al(III) and Fe(III) nitrate. The elution behavior of Pu(IV) was also studied using nitric acid solutions containing reducing agents. More than 90% of plutonium could be recovered from nitric-oxalic acid solutions.     

Solid-phase extraction of plutonium in various oxidation states from simulated groundwater using N-benzoylphenylhydroxylamine

Solid-phase extraction of plutonium in different individual and mixed oxidation states from simulated groundwater (pH 8.5) was studied. The extraction of plutonium species was carried out in a dynamic mode using DIAPAK C16 cartridges modified by N-benzoylphenylhydroxylamine (BPHA). It was shown that the extent of recovery depends on the oxidation state of plutonium. The extraction of Pu(IV) was at the level of 98–99% regardless of the volume and flow-rate of the sample solution. Pu(V) was extracted by 90–95% and 75–80% from 10- and 100-mL aliquots of the samples, respectively, whereas the extraction of Pu(VI) did not exceed 45–50%. An equimolar mixture of Pu(IV), Pu(V), and Pu(VI) was extracted by 74%. The distribution coefficients (K _d) and kinetic exchange capacities (S) of plutonium in various oxidation states were measured. It was found that during the sorption process, Pu(V) was reduced to Pu(IV) by 80–90% after an hour-long contact with the solid phase. Pu(VI) is reduced to Pu(V) by 34% and to Pu(IV) by 55%. In the case of mixed-valent solution of plutonium, only Pu(V) and Pu(IV) were found in the effluents.     

Studies on decontamination of spent ion exchange resin used for plutonium purification in PUREX stream

This paper deals with the studies on decontaminations of spent ion exchange resin used for purification of plutonium in PUREX process stream. Studies were carried out to optimize the chemical procedure for removal of plutonium and fission products activities form spent Ion Exchange resin. Different metal complexing reagents were tested for leaching out of radionuclides entrapped in irradiated spent ion exchange resin. The experimental results indicate that 0.01 M NaF solution was found the most suitable for removal of plutonium. The mixture of Na₂CO₃ and sodium salt of EDTA solution was found to be better for decontamination of spent ion exchange resin from beta and gamma activities. Optimized mixture of 0.5 M Na₂CO₃ and 0.1 M sodium salt of EDTA solution was found to be the most effective for fission product activities removal. After successive multiple contacts using these suitable reagents, the Pu and fission product activities in spent ion exchange resin were brought down to a minimum possible level, making it quite suitable for its long term storage.     

Ion exchange behavior of plutonium(IV) on imidazolium nitrate immobilized resin

Imidazolium nitrate anchored on poly(styrene-divinylbenzene) co-polymer, Im-NO₃, has been synthesized and evaluated for plutonium purification. The results are compared with those obtained using Dowex 1 × 4 anion exchange resin. The distribution coefficient (K_d) of Pu(IV) increased with increase in concentration of nitric acid, reached a maximum at 8 M, followed by decrease in K_d values. Rapid ion exchange of Pu(IV) followed by the establishment of equilibrium occurred within 100 min of equilibration and the data was fitted in to first order rate equation. Variation of distribution coefficient of Pu(IV) as a function of exchange capacity and nitrate ion concentration suggest the involvement of anion exchange mechanism is responsible for extraction. The apparent ion exchange capacity was 310 mg/g at 8 M nitric acid. The performance of the Im-NO₃ under dynamic condition was assessed by column breakthrough experiments. Radiolytic degradation of Im-NO₃ resin in presence and absence of nitric acid (8 M) was studied and the results are reported in this paper.     

Some studies on the extraction of plutonium from phosphate containing nitric acid solutions using DBDECMP as extractant

Extraction studies have been carried out to explore the feasibility of separation of Pu(IV) from phosphate containing analytical wastes generated in the laboratory. Distribution data on the extraction of Pu(IV) from DBDECMP (di-butyl, N,N-diethylcarbamoyl methyl phosphonate) in xylene from an aqueous nitric acid and its mixture with sulfuric as well as with sulfuric and phosphoric acids were obtained. Based on the data obtained, the conditions for the recovery of plutonium from such waste solutions are suggested.     

The behaviour of plutonium in aqueous basic media

Behaviour of Pu(IV) and Pu(VI) in basic media has been investigated by studying their stabilities and quantitative determination by spectrophotometry. Beer's law was found to be obeyed in the range of 1·10^–3 to 5·10^–3 M Pu(IV) at 485 nm peak with a molar absorption coefficient of 95M^–1· cm^–1 in sodium carbonate medium. In case of Pu(VI), in the same medium Beer's law was obeyed in the concentration range of 2·10^–3 to 1·10^–2M at 550 nm with a molar absorption coefficient of 50M^–1·cm^–1. Distribution ratios of Pu(IV) and Pu(VI) for their sorption on Al₂O₃ and Amberlyst A-26 (MP) resin from bicarbonate and carbonate media have been determined. High distribution ratios obtained indicate the feasibility of decreasing the plutonium content of basic carbonate streams in reprocessing. 10% breakthrough capacities for Pu(IV) and Pu(VI) with these exchangers during column operations have also been determined.     

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.     

A new method of plutonium speciation in sea water

A new method of plutonium speciation in large volume of sea water was developed by using adsorption of Pu(IV)-Xylenol Orange chelate and Pu-Arsenazo chelate on XAD-2 resin, respectively. The tetravalent plutonium ion reacts selectively with Xylenol Orange in acid solution and that adsorbed on XAD-2 resin. Total plutonium can be collected onto the resin in the form of its Arsenazo-III complex. The determination of plutonium then was carried out by alpha-ray spectrometric method after decomposition of organic complexes and ion exchange separation. The present method is confirmed for convenient and rapid preconcentration procedure for plutonium shipboard chemistry.     

Influence of manganese dioxide on the oxidation states of plutonium

When manganese dioxide impregnated filters have been used to concentrate plutonium from water solutions some anomalies were detected, which were ascribed to the probable existence of plutonium in two different oxidation states. The results of this paper seem to confirm this assumption that the Pu in the solution is a mixture of Pu(III) and Pu(IV). After contact with MnO₂, plutonium in the solution consists of only Pu(IV).     

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.     

Studies on the coulometric determination of uranium and plutonium employing a graphite electrode

Summary Graphite has been employed as a working electrode in the controlled potential coulometric determination of uranium and plutonium. The couples U(VI)/U(IV) and Pu(IV)/Pu(III) employed for analysis have diverse redox potentials and commonly the working electrodes employed are mercury and platinum. A graphite electrode in the shape of a beaker showed satisfactory performance for the quantitative reduction of U(VI) to U(IV) and Pu(IV) to Pu(III) and also for quantitative oxidation of Pu(III) to Pu(IV). Studies on the levels of the background current, blank values and their reproducible behaviour in acid media have been carried out with a view to achieve good precision and accuracy. A software-based predictive evaluation technique of end-point charge has been investigated. The results have shown that the graphite electrode can be used for the determination of both uranium and plutonium in the presence of each other with a precision and accuracy of better than ±0.5%.     

CARBEX process,a new technology of reprocessing of spent nuclear fuel

Advances in the CARBEX process, a new aqueous chemical method for reprocessing of spent nuclear fuel (SNF) in carbonate media, are considered. A review of carbonate methods for SNF reprocessing is given. The CARBEX process concept is presented and experimental data for every stage of the CARBEX process: high-temperature oxidation of spent fuel composition, its oxidative dissolution in carbonate aqueous solutions, extraction refining of U(VI) and Pu(VI), solid-phase re-extraction of carbonate complexes of U(VI) and Pu(VI), and obtaining of uranium and plutonium dioxide powders for fabrication of ceramic nuclear fuel, are discussed. It was shown that the CARBEX process can be more effective and safe than the well-known industrial PUREX process.     

Determination of plutonium by two-step flow-coulometry at the column electrode

A two-step flow-coulometry method has been developed for rapid determination of elements (plutonium, iron, etc) which exist in various oxidation states in solution, and applied to the determination of plutonium in 0.5M sulphuric acid medium. The first-step column electrode potential is fixed at between +0.10 and +0.35 V vs. Ag-AgCl, and all plutonium ions are reduced to Pu(III). The second-step column electrode potential is fixed at +0.75 V vs. Ag-AgCl, and Pu(III) which flows from the first column electrode is oxidized to Pu(IV). The quantity of plutonium is determined from the number of coulombs used in the oxidation. It is possible to eliminate interference by diverse ions by electroanalysis at the first column electrode. About a 10-mul sample is necessary and the electrolysis for determination is finished in 1 min.