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.     

Rapid separation and purification of uranium and plutonium from dilute-matrix samples

This work describes a streamlined approach to the separation and purification of trace uranium and plutonium in environmental swipe samples that contain a small amount of collected bulk material. We describe key modifications to conventional techniques that result in a relatively rapid, safe, cost-effective, and efficient U and Pu separation process. Simulated samples were produced by loading appropriate ²³⁵U, ²³⁸U, and ²⁴⁰Pu onto high purity cotton swipes. Uranium concentration and isotopic composition were measured by multi-collector inductively coupled mass spectrometry. Corresponding plutonium measurements were conducted with a three stage thermal ionization mass spectrometer. Quantitative U and Pu recoveries were observed with this method.     

Studies on sorption of plutonium from carbonate medium on polyacrylhydroxamic acid resin

Polyacrylhydroxamic acid resin synthesized by functionalization of polyacrylamide with hydroxylamine has been investigated for the sorption of plutonium(IV) from carbonate medium, aiming at its application for the removal of plutonium from alkali wash effluent generated during purification of TBP in PUREX process. Batch experiments have been carried out to determine distribution coefficient of plutonium(IV) between this exchanger and various compositions of carbonate medium. Effect of the concentration of sodium carbonate, sodium bicarbonate and pH of the solution on the distribution coefficient have been studied to optimize the conditions for the uptake of Pu(IV) by this exchanger. Column experiments were carried out to determine the practical capacity of the exchanger for plutonium. Elution studies were also carried out to recover the loaded plutonium from the ion exchange column The exchanger displayed good exchange capacity for Pu(IV) from feed solution simulating the conditions of carbonate wash effluent generated in PUREX process. The exchanger also exhibited fast elution of Pu, suggesting the feasibility of using it for the recovery of Pu from carbonate based wash effluent.     

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.     

Studies on plutonium(IV) hydroxosuccinate

The precipitation behaviour of basic plutonium(IV) compounds from dilute nitric acid medium with aliphatic dicarboxylic acids has been investigated. Unlike plutonium(IV) oxalate, which is precipitated in acid medium (1–4M), the higher dicarboxylates precipitate in 0.9–2.5 pH range and their elemental analysis indicate Pu/dicarboxylate ratio of 11. The mode of formation, composition, solubility and thermal degradation behaviour for plutonium(IV) hydroxosuccinate has been studied. The hydroxy group determination by kinetic titration using fluoride complexing revealed that the compound has a Pu:OH ratio of 11 and determination of bridging oxygen group gave a Pu:O ratio of 10.5 suggest its formation as hydroxysuccinate having an oxobridged formula.     

Electrode processes of plutonium ions in phosphate media

Summary The electrolytic behaviour of plutonium ions in a mixture of phosphate and nitrate solutions was studied by flow-coulometry with column electrodes of glassy carbon(GC)-fibers and voltammetry with a GC disc electrode, and compared with that in phosphate-free media. The redox processes, PuO ₂ ²⁺ /PuO ₂ ⁺ and Pu⁴⁺/Pu³⁺, were demonstrated to be reversible even in the phosphate media and their half wave potentials shifted more negatively due to the formation of PuO₂(H₂PO₄)⁺ and Pu(HPO₄)₂. The rate of the irreversible reduction of PuO ₂ ⁺ to Pu³⁺ increased in the presence of phosphoric acid and the quantitative reduction was attained with the column electrode even at +0.35 V vs. saturated KCl-Ag/AgCl. The reduction process of PuO ₂ ⁺ was elucidated considering an intermediate Pu(IV)-species, PuO²⁺, which decomposed into Pu⁴⁺ by a post-chemical reaction. Analytical advantages of the use of phosphate media are discussed.     

Studies on the age determination of trace plutonium

An isotope dilution multicollector inductive coupled plasma mass spectrometry (ID-MC-ICP-MS) method for determining age of trace Pu through measuring ²⁴¹Pu/²⁴¹Am, ²⁴⁰Pu/²³⁶U ratio was established. At the same time, other two methods-α-spectrometry combined with MC-ICP-MS and liquid scintillator combined with α-spectrometry through measuring ²⁴¹Pu/²⁴¹Am ratio to determine the age of trace Pu were also studied. The techniques were explored for the age determination of nanogram grade Pu sample on the basis of Pu/Am, Pu/U separation. The ages of two Pu samples—one with known and the other with unknown age—were determined by the three methods. The determined ages by the three methods were all in agreement with the reference value. The established methods for determining the age of trace Pu could be adopted in the verification activities of nuclear safeguards and nuclear arms control.     

Selective sensor for silver ions built from polyfluorophores on a DNA backbone

To explore a new modular metal ion sensor design strategy, fluorophores and ligands were incorporated into short DNA-like oligomers. Compound 1 was found to function as a selective sensor for Ag(+) in aqueous buffer, where low micromolar concentrations of Ag(+) induce a red-shifted, turn-on fluorescence signal. Experiments with HeLa cells show that 1 can penetrate cells and yield a signal for intracellular Ag(+). This suggests a broadly applicable approach to developing sensors for a wide variety of cations.     

Recovery of americium from analytical solid waste containing large amounts of uranium,plutonium and silver

Trace metallic impurity analysis by spectroscopic techniques is one of the important steps of chemical quality control of nuclear fuel materials. Depending on the burn-up and the storage time of the fuel, there is an accumulation of ²⁴¹Am in plutonium based fuel materials due to β decay of ²⁴¹Pu. In this paper, attempts were made to develop a method for separation of ²⁴¹Am from 1.2 kg of analytical solid waste containing 70% U, 23% Pu, 5% Ag and 1–2% C as major constituents along with other minor constituents generated during trace metal assay of plutonium based fuel samples by d. c. arc carrier distillation atomic emission spectrometry. A combination of ion exchange, solvent extraction and precipitation methods were carried out to separate ~45 mg of ²⁴¹Am as Am(NO₃)₃ from 15 L of the analytical waste solution. Dowex 1×4 ion exchange chromatographic method was used for separation of Pu whereas 30% TBP–kerosene was utilized for separation of U. Am was separated from other impurities by fluoride precipitation followed by conversion to nitrate. The recovery of Pu from ion exchange chromatographic separation step was ~93% while the cumulative recovery of Am after separation process was found to be ~90%.     

Studies on the adsorption of plutonium (IV) on alumina from phosphoric acid-nitric acid solution

Results of experiments on the adsorption of plutonium (IV) on alumina and their application to the recovery of plutonium from analytical waste solutions containing phosphoric-nitric acid are reported. Distribution ratios of plutonium (IV) between alumina and solutions containing varying concentrations of phosphoric acid and nitric acid are determined. The influence of various ions like UO₂ ²⁺, Fe³⁺, MoO₄ ^2–, VO²⁺ and SO₄ ^2– on the distribution ratio is evaluated. Saturation values of adsorption of plutonium (IV) on alumina and optimum conditions for loading and elution of plutonium on a column packed with alumina are described.     

Recovery and purification of plutonium in lean plutonium product stream obtained during reprocessing of fast reactor spent fuels

A method has been developed for final purification of plutonium from uranium and fission products of high beta gamma activity. This method involves selection of a suitable ion exchange resin for the purification of plutonium in order to deliver a quality PuO₂ product. The effect of the concentration of uranium and plutonium, effect of increased loading of uranium and number of bed volumes for effective washing, which are some of the parameters that generally affect the recovery and purification of plutonium were investigated. An excellent decontamination factor for fission products has been achieved by this anion exchange process which in turn delivered an excellent PuO₂ product quality in terms of purity and associated beta gamma activity with low personnel radiation exposure.     

Studies on reductive back extraction of plutonium in Purex process

Kinetic equation of one stage process of reduction of plutonium(IV) in two phase 30 vol. % TBP — n-dodecane/HNO₃ system is given. The expression derived is used for elaborating the method of calculation of equilibrium (ultimate) concentration profiles of plutonium in multistage contactors for countercurrent process of U/Pu separation.     

Sorption of traces of silver ions onto polyurethane foam from acidic solution

The sorption of traces of silver ions onto polyurethane foam (PUF) has been investigated in detail. Maximum sorption of silver (K(d)=6109 cm(3) g(-1), %sorption>97.5%) has been achieved from 1 M nitric acid solution after equilibrating silver ions with approximately 29 mg PUF for 20 min. The kinetics and thermodynamics of the sorption of silver ions onto PUF have also been studied. The sorption of silver ions onto PUF follows a first-order rate equation, which results as 0.177 min(-1). The variation of sorption with temperature yields the values of DeltaH=-56.1+/-3.2 kJ mol(-1), DeltaS=-159.7+/-10.5 J mol(-1) K(-1) and DeltaG=-8.68+/-0.09 kJ mol(-1) at 298 K with a correlation factor gamma=0.9919. The sorption data were subjected to different sorption isotherms. The sorption follows Langmuir, Freundlich and Dubinin-Radushkevich (D-R) isotherms. The values of Langmuir isotherms Q=65.4+/-1.5 mumol g(-1) and b=(4.79+/-1.16)x10(4) dm(3) mol(-1) have been evaluated for Langmuir sorption constants, whereas the Freundlich sorption isotherm gives the value 1/n=0.12+/-0.02 and A=0.15+/-0.03 mmol g(-1). The D-R parameters computed were beta=-0.000817+/-0.000206 mol(2) kJ(-2), X(m)=76.8+/-8.7 mumol g(-1) and E=24.7+/-3.2 kJ mol(-1). The influence of common ions on the sorption was also examined. It is observed that Hg(II), thiourea, Al(III), thiocyanate and thiosulphate reduce the sorption, whereas Cu(II), citrate and acetate ions enhance the sorption significantly. It can be concluded that PUF may be used to remove traces of silver ions from its very dilute solutions or for its preconcentration from aqueous acidic solutions.     

Studies on the extraction behaviour of plutonium(IV) from mixed aqueous-organic media with sulphoxides

Extraction behaviour of plutonium (IV) from nitric acid media by two long-chain aliphatic sulphoxides, namely, di-n-hexylsulphoxide and di-n-octylsulphoxide has been investigated in the presence of several water-miscible organic solvents to study their possible synergistic effect on metal ion extraction. Methanol, ethanol, n-and iso-propanol, dioxane, acetone as well as as acetonitrile were used as the organic component of the mixed (polar) phase. These additives affected the extraction to varying degrees. Thus, extractability of Pu increases 2–3 fold with increasing concentration (upto 20%) of acetonitrile, acetone, methanol and ethanol while it decreases with increasing concentration of n-and isopropanol. At high concentration of the former, synergism changes into antagonism. Possible reasons for such behaviour are briefly discussed. Among these organic additives, maximum enhancement in the extraction of Pu(IV) was observed in the presence of acetonitrile. The relative increase in extraction was found to be more at lower sulphoxide concentrations.     

银纳米粒子的形貌可调控研究

在聚乙烯吡咯烷酮(PVP)保护下,以乙二醇(EG)为还原剂制备银纳米粒子.探讨了反应物浓度、反应温度对制备的纳米银粒子形貌的影响.采用X射线衍射(XRD)和透射电镜(TEM)来表征纳米银粒子的结构和形貌.结果表明,AgNO3和PVP的浓度,AgNO3和PVP的比例以及反应温度对纳米银粒子有较大影响,反应温度控制在140 ℃至160 ℃之间,易于控制纳米银粒子的形貌.     

Optode membrane for continuous measurement of silver ions

An optical sensor (optode) membrane for continuous monitoring of silver ion is presented. It is composed of plasticized poly(vinyl chloride) (PVC), a silver-sensitive chromo-ionophore, and a lipophilic borate salt acting as an anionic site. The membrane selectively responds to silver ion in the concentration range from 50 M to 10 mM, giving a reversible colour change from blue to pink, with response times of the order of 2 min. The acridinium dye acts as both the recognition element and the chromo—ionophore. Its long-wave absorption band is at around 680 nm, which makes it compatible with existing LED light sources and diode lasers. A lipophilic octadecyl side chain renders it highly lipophilic and this prevents its leakage out of the PVC membrane. Hence, the operational lifetime exceeds 1 week, without any changes in the work function.     

Preventing pollution from plutonium processing

The plutonium processing facility at Los Alamos has adopted the strategic goal of becoming a facility that processes plutonium in a way that produces only environmentally benign waste streams. Pollution prevention through source reduction and environmentally sound recycling are being pursued. General approaches to waste reductions are administrative controls, modification of process technologies, and additional waste polishing. Recycling of waste materials, such as spent acids and salts, are technical possibilities and are being pursued to accomplish additional waste reduction. Liquid waste stream polishing to remove final traces of plutonium and hazardous chemical constituents is accomplished through (a) process modifications, (b) use of alternative chemicals and sorbents for residue removal, (c) acid recycling, and (d) judicious use of a variety of waste polishing technologies. Technologies that show promise in waste minimization and pollution prevention are identified. Working toward this goal of pollution prevention is a worthwhile endeavor, not only for Los Alamos, but for the Nuclear Complex of the future.     

Long time contamination from plutonium

The long-term effect of the plutonium to the human organism after Chernobyl accident was traced. Nine years after the accident the level of plutonium in the urine of the patients was in the order of tenths mBq.