Ceric tungstate as sorbent for polyvalent tracer cations and as an ion exchanger for separation of carrier-free95Nb from95Zr and113mIn from113Sn

The uptake of a few polyvalent ions Ca²⁺ Cu²⁺, Zn²⁺, Cd²⁺, UO²⁺, Cr³⁺, Y³⁺, Ce³⁺, Nd³⁺, Sm³⁺, Tb³⁺, Tm³⁺, Yb³⁺, Lu³⁺, Zr⁴⁺, Hf⁴⁺, Sn⁴⁺, Nb⁵⁺, Se⁶⁺, Mo⁶⁺ and W⁶⁺ at very small concentrations has been studied over ceric tungstate exchanger. A good column variety of the material was prepared by mixing ceric sulphate in 2N H₂SO₄ and aqueous solution of sodium tungstate in suitable proportion. The ratio of cerium: tungstate was obtained to be 11. Separation of carrier-free⁹⁵Nb from⁹⁵Zr and^113mIn from¹¹³Sn have been carried out by applying a very simple chemical procedure over the column of ceric tungstate. The -spectrum of separated⁹⁵Nb and^113mIn products were found to be of high radionuclidic purity. The separation procedure took less than 15 min and the yields were close to 100%.     

Separation of113mIn from113Sn on an isotope generator with inorganic ionite

A method of preparation of hydrated zirconium oxide suitable for^113mIn generators was elaborated. A good separation of^113mIn from¹¹³Sn was obtained in the course of routine use of generator, with a very small admixture of zirconium in the eluate.     

Separation of radioindium (N.C.A.) from neutron irradiated tin and cadmium on hydrated antimony pentoxide

Hydrated antimony pentoxide (HAP) as an absorbent for column operation has been prepared by hydrolysis of SbCl₅ with deionized water. Sorption behavior of Sn, Cd, Sb and In was studied on HAP in HCl medium. Radiochemical separation of no-carrier added^113mIn from¹¹³Sn/¹²⁵Sb and^115mIn from¹¹⁵Cd wa achieved over a HAPO column. The separated products were of high radionuclidic purity.     

Synthesis of radioactive tributyl[113Sn]tin of high specific activity for use in environmental fate studies

A method of synthesis of tributyl[¹¹³Sn]tin,­(n/C₄H₉)₃¹¹³Sn(IV), from commercially avail-­able inorganic ¹¹³Sn(IV) is presented. Inorganic tin is first extracted in diethyl ether and reacted with C₄H₉MgCl to produce tetrabutyltin, (C₄H₉)₄¹¹³Sn, which is then debutylated with HgCl₂. The resulting tributyl[¹¹³Sn]tin chloride is isolated from the reaction mixture by successive extractions with hexane and aqueous Na₂S₂O₃. The yield is 40–60% and the product obtained is >98% pure. It has the same specific activity as the starting ¹¹³Sn(IV), i.e. up to 550 MBq mg⁻¹ Sn, making it suitable for use in environmental fate and toxicology studies at concentrations relevant of those found in the aquatic environment. © 1998 John Wiley & Sons, Ltd.     

Separation of trace elements,In(III), Sn(IV), Sb(V) and Te(IV) by adsorption on activated carbon and graphite

Adsorption behaviour of trace elements, In(III), Sn(IV), Sb(V) and Te(IV) on activated carbon and graphite powder was studied. Adsorption characteristics of the ions enabled the separation of In(III)–Sn(IV), Sn(IV)–Sb(V) and Sb(V)–Te(IV) pairs. Applications to practical separation, milking of^113mIn from¹¹³Sn, removal of tin impurity from¹¹⁹Sb, and milking of¹¹⁹Sb from^119mTe, are presented.     

Separation of <Superscript>113m</Superscript>In from <Superscript>113</Superscript>Sn based on activated carbon used as column matrix

The activated carbon was prepared by using corncobs and characterized by sorpatometer for using as an exchanger material to separate the generated ^113mIn from ¹¹³Sn and ^124,125Sb. To optimize the separation process, the different parameters like acetone percentage, HCl concentration were studied. The exchange capacity of Sn(IV) is 7.6 meq/g onto the activated carbon and the elution efficiency of ^113mIn > 80% by using 10 mL of 0.2 M HCl-80% acetone with flow rate 1 mL/min. The radionuclidic purity and radiochemical purity of the eluted ^113mIn were examined and clarified the presence of ^124,125Sb with relatively high level as radio impurities, so further separation was carried out by using Dowex 1×8 as an anion exchanger below the activated carbon matrix on the same separation column to adsorb the ¹¹³Sn and ^124,125Sb, which escape from the activated carbon matrix.     

Removal of 63Ni and 57Co from aqueous solution using antimony doped tin dioxide–polyacrylonitrile (Sb doped SnO2–PAN) composite ion-exchangers

Tin dioxide and its antimony doped counterpart were synthesized using traditional sol–gel procedure. The metal oxides were then turned into composites by mixing them with polyacrylonitrile (PAN) and composite spheres ready for use in traditional column applications were obtained. The characterization of materials was investigated by X-ray diffraction, scanning electron microscopy–energy dispersive X-ray, surface area, point of zero charge and thermal analyses. Static batch experiments showed that the antimony doped tin dioxide–PAN (Sb doped SnO₂–PAN) is an effective material for nickel removal and the composite maintains its good metal uptake properties in dynamic column conditions. The composite showed a high nickel uptake capacity of 9 mmol/g in 0.1 M NaNO₃ solution. It was observed that the ion exchange kinetics of antimony doped tin dioxide (Sb doped SnO₂) was remarkably fast for ⁵⁷Co and ⁶³Ni ions but turning the material into PAN composite significantly decreased the materials kinetic properties.     

Ga2O for target, solvent extraction for radiochemical separation and SnO2for the preparation of a68Ge/68Ga generator

Summary The target for the production of⁶⁸Geconsists of a disc of gallium suboxide, Ga₂O, with a 19 mm diameter. The suboxide was primarily prepared by repeatedly mixing metallic Ga and Ga₂O₃at 700 °C. The target (2.4 g) was quite stable under a long-time irradiation with a 34 MeV proton beam at a current of ~80<span lang=EN-ZA style='font-size:12.0pt; font-family:Symbol;mso-bidi-font-family:Symbol;mso-ansi-language:EN-ZA'>mA. The dissolution of the target was performed using 12M sulphuric acid solution, assisted with the dropwise addition of 30% H₂O₂solution, and took less than 4 hours. A solvent extraction method, using a 9M H₂SO₄-0.3M HCl/CCl₄system, was employed for the radiochemical separation of⁶⁸Ge from Ga and Zn radionuclides, while 0.05M HCl was used for the back extraction of⁶⁸Ge from the organic phase. The⁶⁸Ge obtained in the dilute HCl was directly loaded onto a column containing either a hydrous tin dioxide or a crystalline tin dioxide, obtained by calcinations of the hydrous oxide at 450, 700, and 900 °C. The calcinated hydrous tin dioxide at 900 °C showed the highest crystallinity and highest⁶⁸Ga elution yield and was selected for use in the generator. The⁶⁸Ga elution from the column generator packed with 2 g of tin dioxide, using 3 ml of 1M HCl, and yielded an average of 65%. The breakthrough of⁶⁸Ge was 6.1^. 10^-4%.     

Separation of tin from some metals by extraction chromatography

Much attention has been devoted to Sn (IV) strongly retained on the TBP-Daiflon column from 2M HCl in extraction chromatography. The separations of Sn?Cu, Zn, As, Cd, Sb, Pb,¹¹³Sn-¹²⁵Sb,¹¹³Sn-^113mIn (^113mIn milking) and Sn?Hg?Fe were successfully achieved without any contamination. In the separations, except for the last, only tin was retained separately on the column upon passing the mixed solution. The daughter indium was eluted with 0.5M HCl. In the last separation, iron was eluted with 0.5 M HCl, tin with 0.1M HCl and mercury with 2M HNO₃, for these metals retained on the column. Radioactive tracers for tin, iron, mercury and antimony were used.     

Zirconium silicotungstate matrix as a prospective sorbent material for the preparation of <Superscript>113</Superscript>Sn/<Superscript>113m</Superscript>In generator

The zirconium silicotungstate (ZrSiW) was studied as an effective sorbent material to be used in the ¹¹³Sn/^113mIn generator. The results elucidated that the distribution coefficient of ¹¹³Sn (3700 mL/g) is greater than ^113mIn (275 mL/g) from 0.1 M HCl acid solution to the ZrSiW material. The maximum sorption capacity of Sn (IV) was found to be 33 mg per gram ZrSiW (~?0.3 mmol/g). The elution yield of ^113mIn was found to be >?78?±?6.4% with an acceptable purity of radionuclidic and radiochemical (≥?99.99 and 96.8%, respectively). The rigorous separation of ^113mIn from the ¹²⁵Sb was carried out due to its long half-life (2.758 years) and beta emission that causes tissue damage. Zr, W and Si levels are below the permitted limit in the ^113mIn eluate.     

Potential use of tungstocerate matrices on the separation of <Superscript>113m</Superscript>In from <Superscript>110m</Superscript>Ag relevant to the separation of the cyclotron-produced <Superscript>111</Superscript>In from its silver target

A procedure for separation of no-carrier-added ^113mIn(III) radioisotope from a bulk of ^110mAg has been developed. The sorption behavior of ^113mIn(III) and ^110mAg(I) ions in HNO₃ acid solutions on different tungstocerate matrices showed high affinity of ^110mAg(I) ions towards tungstocerate(IV) gel matrices compared with ^113mIn(III) ions. No-carrier-added ^113mIn radionuclide was separated from ^110mAg on 12-tungstocerate(IV) column matrix. 11 mL 0.3M HNO₃ acid solution was enough for eluting the ^113mIn from the column bed. ^110mAg was recovered from the column by eluting the column bed with 12 mL 2M HNO₃ acid solution.     

Dünnschichtchromatographie als Hilfsmittel in der Radiochemie. 4. Mitteilung Trennung von 113Sn und 113mIn

Thin-layer chromatography is applied to the separation of ^113mIn from ¹¹³Sn. Different factors influencing migration and separation of the ions are investigated. The method can be used for the examination of the purity of isolated ^113mIn solutions.     

Separation of Ba and Pb from aqueous solutions of Sr with commercially available “precipitated active manganese dioxide”

Ba and Pb radionuclides can be removed from Sr in aqueous solution in both sodium acetate and acetic acid, containing 20 mg Sr carrier, by stirring with small (0.1–0.5 g) amounts of solid manganese dioxide. In tracer experiments⁸⁵Sr was separated with only small losses from¹³³Ba and²¹⁰Pb by separation factors of 87 and 135, respectively. The separation factor is defined here as the % of the initial⁸⁵Sr activity/ % of the initial¹³³Ba or²¹⁰Pb activity remaining in the aqueous phase after MnO₂ contact. The applicability of this technique for removing Ba and Pb radionuclides in the analysis of⁹⁰Sr in environmental samples (especially milk) is discussed.     

Hydrogen production from bio-oil by chemical looping reforming

Hydrogen is a green energy carrier. Chemical looping reforming of biomass and its derivatives is a promising way for hydrogen production. However, the removal of carbon dioxide is costly and inefficient with the traditional chemical absorption methods. The objective of this article is to find a new material with low energy consumption and high capacity for carbon dioxide storage. A metal organic framework (MOF) material (e.g., CuBTC) was prepared using the hydrothermal synthesis method. The synthesized material was characterized by X-ray diffraction, ?196 °C N₂ adsorption/desorption isotherms, and thermogravimetry analysis to obtain its physical properties. Then BET, t-plot, and density functional theory (DFT) methods were used to acquire its specific surface area and pore textural properties. Its carbon dioxide adsorption capacity was evaluated using a micromeritics ASAP 2000 instrument. The results show that the decomposition temperature of the synthesized CuBTC material is 300 °C. Besides, high CO₂ adsorption capacity (4 mmol g^?1) and low N₂ adsorption capacity were obtained at 0 °C and atmospheric pressure. These results indicate that the synthesized MOF material has a high efficiency for CO₂ separation. From this study, it is expected that this MOF material could be used in adsorption and separation of carbon dioxide in chemical looping reforming process for hydrogen production in the near future.     

Comparative study of different chromatographic methods for quality control of113mIn-radiopharmaceuticals

Paper chromatography and Gel Chromatography Column Scanning technique (GCS) have been applied for the separation of indium fractions in^113mIn-radiopharmaceuticals. By these techniques the percentage of ionic indium,^113mIn-colloid and^113mIn-compound have been determined. The resolution efficiency of the gels was found to be significantly influenced by the gel type media and the pH of the eluent. The results obtained from the GCS-profiles indicated that the Sephadex G-50 Fine was the best and can be routinely used in the radiochemical quality control of the^113mIn-phytate. Good separation of^113mIn-colloids,^113mIn-microaggregate and^113mIn-phytate from carrier-free^113mIn-eluate was performed using Whatman No. 3, previously washed with 0.04N HCl and developed either with 0.9% NaCl (for^113mIn-colloid), or 85% methanol (for^113mIn-phytate), or phosphate-methanol buffer (for^113mIn-microaggregate) as rapid and simple procedure for determination of the radiochemical quality control of indium compound in the forms of radiocolloids.     

Separation of carrier-free115mIn from115Cd and132I from132Te over the zirconium oxide column

A new way of radiochemical separation of carrier-free^115mIn from¹¹⁵Cd and¹³²I from¹³²Te over the column of zirconium oxide is described. Activities of Cd and In in equilibrium in dilute acetic acid solution were bufferred with dilute sodium acetate and fed into the column at a pH 7, when cadmium activity passed out unadsorbed and the^115mIn was adsorbed in the column. A study of the γ-ray peak of the separated^115mIn showed that the product is of high radionuclidic purity. Te-I pair was separated by passing the weakly acidic solution of¹³²Te and¹³²I in the presence of AgNO₃ and Na₂SO₃, through the column where both activities were adsorbed. Iodine was washed outh with 5% sodium thiosulphate solution and the retained tellurium activity was later washed out with 6M HNO₃. The β-decay study showed that the separated¹³²I product is of high radiochemical purity. The processes took less than half an hour and the yields were quantitative.     

Reuse of decayed tellurium dioxide target for production of iodine-131

Iodine-131 was generated by irradiation of natural tellurium dioxide in a nuclear reactor. After irradiation the tellurium dioxide was transferred to hot cell and heated in a quartz furnace at ~700 °C. The Iodine-131 was distilled and collected in carbonate/bicarbonate buffer and used for thyroid cancer patients. The tellurium dioxide used was >99 % pure. During nuclear reaction nanogram tellurium was consumed to produce ¹³¹I, although significant loss of target material may occur in heating process. In dry distillation technique no chemical was added to irradiated target material. After ¹³¹I separation tellurium dioxide was decayed for 8 years in radioactive waste management facility. The decayed TeO₂ was recovered, melted and crushed for desired mesh size. The TeO₂ was sealed in Al capsule for re-irradiation purposes. The separation of ¹³¹I was carried out via dry distillation. The purity and yield of ¹³¹I separated from both of the irradiated new and irradiated re-used tellurium dioxide targets were comparable.     

Improved processes of molybdenum-99 production

Two improved processes of⁹⁹Mo production have been developed on laboratory scale. The first one allows to purify Mo of natural isotopic composition from tungsten impurities from 64 to <10 ppm by using preferential adsorption of tungsten on hydrated tin(IV) oxide (SnO₂ nH₂O) before irradiation in a nuclear reactor. The second process deals with the separation of pure fission product⁹⁹Mo from²³⁵U irradiated in a reactor. Two versions of separation process for production of fission⁹⁹Mo have been developed. Both versions start with the dissolution of²³⁵U oxide target in nitric acid and are based on sequential use of alumina and anion exchange resin AG® 1-X8 columns. The yield of⁹⁹Mo in both versions is 80–89%.     

Study of tin dioxide–sodium stannate composite obtained by decomposition of peroxostannate as a potential anode material for lithium-ion batteries

A tin dioxide–sodium stannate composite has been obtained by the thermal treatment of sodium peroxostannate nanoparticles at 500°C in air. X-ray powder diffraction study has revealed that the composite includes crystalline phases of cassiterite SnO₂, sodium stannate Na₂Sn₂O₅, and sodium hexahydroxostannate Na₂Sn(OH)₆. Scanning electron microscopy has shown that material morphology does not change considerably as compared with the initial tin peroxo compound. Electrochemical characteristics have been compared for the anodes of lithium-ion batteries based on tin dioxide–sodium stannate composite and anodes based on a material manufactured by the thermal treatment of graphene oxide–tin dioxide–sodium stannate composite at 500°C in air.     

Elution dynamics of adsorption of Chladon 113 and methyl bromide on active carbons

The elution dynamics of adsorption of Chladon 113 (at high contents of the adsorptive) and methyl bromide on active carbons with different lengths of the layer was studied. The adsorption isotherms were described by the equation of the theory of volume filling of micropores. The elution curves were calculated using the model of the equilibrium adsorption layer. The effective kinetic constant calculated in the framework of the model is 1.5 cm for Chladon 113 at a linear velocity of vapor-air flow of 1250 cm min⁻¹ and 1.0 cm for methyl bromide at the velocity of 700 cm min⁻¹. Translated fromIzvestiya Akademii Nauk. Seriya Khimicheskaya, No. 7, pp. 1353–1355, July, 1998.