Separation of iodine produced from fission using silver-coated alumina

A method developed for the preparation of silver-coated alumina, a new material for retention of iodine from alkaline solution is described. Experiments were carried out, oriented to the purification step of ⁹⁹Mo produced by uranium fission, based on the retention of radioiodine in this material. Iodine retention, as well as molybdenum non-retention, were found, both with excellent results. Further tests showed that the incorporation of this material has no influence on the subsequent ⁹⁹Mo retention in ion exchange resins. The elution of the radioiodines retained was tested with satisfactory results. This new material can be used not only for improving the ⁹⁹Mo purification and working conditions, but also as the basis of a method for recovering the fission produced ¹³¹I.     

Separation of 99Mo from a simulated fission product solution by precipitation with a-benzoinoxime

The formation property of Mo precipitate was investigated and improved the existing process was using H₂O₂ that acts as an interfering compound in a subsequent alumina adsorption process. The property of the Mo precipitate was investigated by using SEM, FTIR, TG-DTA, and XRD. The simulated solution consisted of 1M nitric acid containing seven elements (Mo, I, Ru, Zr, Ce, Nd, Sr) and their radioactive tracers. As a result, the precipitate was composed of the Mo precipitate and re-precipitated a-benzoinoxime which was added excessively for increasing the precipitation efficiency. It was confirmed that the Mo precipitate was formed by the reaction of two a-benzoinoxime molecules and one MoO₂ ²⁺. Molybdenum precipitate was dissolved in 0.4M NaOH solution within 5 minutes without H₂O₂. Hydrogen peroxide induced only the rapid dissolution of the a-benzoinoxime re-precipitate. Also, the dissolution method without H₂O₂ was favorable in the purification aspect because Zr and Ru were contained as a small fraction of 1.3% and 7.7%, respectively, in the dissolving solution.     

Separation and purification of 99mTc from 99Mo produced by electron linear accelerator

The medical radionuclide ⁹⁹Mo was produced by the ¹⁰⁰Mo(γ,n) reaction using bremsstrahlung photons generated by an electron linear accelerator. The amount of ⁹⁹Mo produced was compared to that predicted by calculation using the particles and heavy ion transport code system. From the ⁹⁹Mo produced, highly pure ^99mTc was separated using the so-called technetium master milker, and the chemical yield of ^99mTc was 83–99 %. The installation of a new complex using this method and the electron linear accelerator with the preferable specification was suggested, and a possibility to supply the demand of ^99mTc was discussed and shown.

     

Separation of99Mo from132Te using thiourea as complexing agent. Application to the separation of99Mo from fission products

A radiochemical method to isolate⁹⁹Mo from¹³²Te, both produced in the fission of²³⁵U, has been developed. The method is based on the formation of a cationic complex of tellurium with thiourea in acid medium which is retained (98.7±0.5)% on a cation exchange resin (Dowex 50W-X8, 100–200 mesh), while (99.8±0.05)%⁹⁹Mo passes through it, due to the non-formation of such complex in the same experimental conditions. The radionuclidic purity of⁹⁹Mo was found to be suitable for the preparation of⁹⁹Mo–^99mTc generators. The retention of⁹⁹Mo on an alumina column as a function of pH was investigated and the best pH range for this purpose was found to be 4.0–4.5.     

Separation of99mTc from parent99Mo by solid-phase column extraction as a simple option for a new99mTc generator concept

Evidence is obtained to show that the liquidliquid extraction separation of^99mTc from⁹⁹Mo with methyl ethyl ketone, methyl propyl ketone and methyl isobutyl ketone can be transformed into a solid-phase column extraction procedure. The aqueous alkaline⁹⁹molybdate solution is immobilized on a column of a granular large-pore diatomaceous earch support, which is the neluted with the abovementioned extractants. Rapid and clean separation of^99mTc can be with all three solvents. The^99mTc can be back-extracted from the organic phase on a column filled with distilled water /or saline/ loaded granular diatomaceous earth /Extrelut^®/. The possibility of using the abovementioned procedure as a basis for a new^99mTc/⁹⁹Mo generator concept is envisaged.     

Separation of trace impurities of131I from99Mo obtained by extraction from fission products of uranium

The radioisotope⁹⁹Mo was separated from a mixture of²³⁵U fission products in the presence of Hg²⁺ by sorption on a chromatographic column filled with Al₂O₃ and elution with 1M NH₄OH. Trace impurities of¹³¹I in the molybdenum fraction were eliminated by selective sorption on silver iodide or by repeated sorption of⁹⁹Mo on Al₂O₃.     

Determination of gamma impurities in fission product99Mo

The methodology is given for determination of very low quantities of gamma impurities:⁹⁵Zr,⁹⁵Nb,¹⁰³Ru,¹³¹I,¹³²Te,¹³²I,¹⁴⁰Ba,¹⁴⁰La,¹⁴¹Ce,¹⁴⁴Ce in⁹⁹Mo produced by separation from other fission products, by means of different measurement conditions such as initial activity, decay time and attenuation with lead.Presented as a poster at the International Conference on Nuclear and Radiochemistry, Beijing, P.R. China, September, 1986.     

Separation of fission produced 106Ru from simulated high level nuclear wastes for production of brachytherapy sources

An effective and simple process for the isolation of ¹⁰⁶Ru from high-level liquid wastes was developed. Radioactive ruthenium was oxidized by H₅IO₆ in HNO₃ solution and was extracted to CCl₄ phase in the form of RuO₄. In order to obtain ruthenium in the suitable form for production of brachytherapy sources, RuO₄ in organic phase was reduced and re-extracted to aqueous phase. The efficiency of extraction of ¹⁰³Ru to organic phase was 86 %, re-extraction to aqueous solution was near 100 %, so the overall recovery of ¹⁰³Ru is estimated at more than 80 %.     

Selective separation of99Mo from fission products in chloride media on activated alumina

The radioisotope⁹⁹Mo generated through the fission of neutron-irradiated uranium targets was separated in an extremely radiochemically pure state. The target dissolution mixture was run on a silica gel column whereby the⁹⁵Zr−⁹⁵Nb activity, the serious radiocontaminants of⁹⁹Mo, was adsorbed on the gel. The effluent from the gel column was processed and treated with sodium dihydrogen phosphate. The processed solution was run on an activated alumina column whereby⁹⁹Mo was adsorbed as phospho-molybdate complex ion, then desorbed from the column as the molybdate ion by eluting with 10% NaOH solution. By recycling the adsorption and desorption steps on alumnia,⁹⁹Mo was finally obtained totally free from all other radiocontaminants.     

Desorption of99Mo from spent99Mo/99mTc generator

A simple method for desorption and purification of⁹⁹Mo from spent⁹⁹Mo/^99mTc generators is described. The alumina column was washed successively with 0.9% saline water, 35% H₂O₂, and then the⁹⁹Mo was eluted with 2M NH₄OH. Ammonia and residual H₂O₂ were removed by heating the eluate. Finally,⁹⁹Mo solution was passed through a 0.2 m membrane filter to remove precipitated aluminium hydroxide.     

99Mo提取后的裂变残液中铀回收技术初步研究

本文对裂变99Mo提取后残液中铀的回收进行了不同类型萃取溶液的效率对比,最终选择了萃取效率最好的30%二-(2-乙基乙基)磷酸的正十二烷溶液进行了条件实验,结果表明体系的HNO3浓度在4.0mol/L附近铀的分配比达到最大值,萃取过程为放热反应,且提高搅拌强度有利于加快萃取速率。     

Estimation of 99Mo production rates from natural molybdenum in research reactors

Journal of Radioanalytical and Nuclear Chemistry - Molybdenum-99 is one of the most important radionuclides for medical diagnostics. In 2015, the International Atomic Energy Agency organized a...     

Fast separations of yttrium,silver, bromine and iodine in heavy-ion-induced fission

Fast separation methods were devised for short-lived fission products embedded in Mg catcher foils; these are based on solvent extraction techniques.³⁴Cl is a major contamination problem because it is formed directly by spallation in the catcher foil. The methods for Ag, Br and I had a decontamination factor of >1000: 1 for this isotope.     

Chemical states of iodine produced by the spontaneous fission of252Cf

The distributions of the chemical states of¹³¹I,¹³²I,¹³³I,¹³⁴I and¹³⁵I produced by the spontaneous fission of²⁵²Cf were investigated by the method of fractional precipitation separation in 6M NH₄OH solution followed by measurement of the X-ray spectrum of the samples with a Ge(Li) detector. The iodine fission fragments were separately collected in the matrices of NaNO₃, NaCl, dehydrated sodium sulfate, CH₃COONa·3H₂O, Na₃H₂IO₆, KI and NaIO₃ crystals. The results showed that the distribution of the chemical states was the function of the collection time, the nature of matrix materials and the type of iodine isotopes. The relative amount of the reduced states of¹³⁴I decreased with the increase of collection time in the matrix of NaNO₃. The ratio of the amount of reduced states to the total amount of the¹³⁴I was the highest (50.4%) in NaCl matrix and the lowest (9.4%) in Na₃H₂IO₆ matrix. The relative quantities of reduced states of¹³⁵I were the highest (58.5% and 97.7%) and that of¹³⁴I were the lowest (16.9% and 63.1%) in matrices of NaNO₃ and NaCl, respectively.     

Characteristics and behavior of a99Mo/99mTc generator using irradiated titanium molybdate as column matrix

A⁹⁹Mo/^99mTc generator, system was made with a performed titanium molybdate gel. The irradiation was carried out at a medium neutron flux of 1.5×10¹³ n cm^–2·s^–1. The irradiated matrix was loaded on top of a column composed of hydrous zirconium oxide alumina. The elution efficiency and the amount of total technetium per mCi^99mTc in the generator eluents have been determined. Molybdenum breakthrough has also been determined and compared with literature values. The influence of the particle size, water content, neutron flux and molybdenum content on the total^99mTc-activity has been investigated.