The determination of molybdenum in geological samples by Neutron Activation Analysis

A metal-silicate extraction technique combined with neutron activation analysis has been developed to determine molybdenum in geological samples. The samples are equilibrated with Femetal powder at high temperatures. Molybdenum is completely extracted into the metal phase because of very reducing conditions in the furnace. The metal spherule is separated from the silicates, irradiated and dissolved in an acid solution. The molybdenum is precipitated as a sulfide and the precipitate is dissolved in aqua regia and counted on a Ge/Li/detector. The radiochemical yield is obtained by irradiation of the solution. The method avoids production of⁹⁹Mo from induced fission of²³⁵U by performing the metal-silicate separation before irradiation. The precipitation step may be necessary to remove the high background from the decay of⁵⁹Fe. Mo concentrations down to 15 ng/g have been obtained using this method.     

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.     

Evaluation of proton induced reactions on 100Mo: New cross sections for production of 99 mTc and 99Mo

The use of the ⁹⁹Mo^99mTc generator in nuclear medicine is well established world wide. The production of the ⁹⁹Mo (T_1/2 = 66 h) parent as a fission product of ²³⁵U is largely based on the use of reactor technology. From the early 1990's accelerator based production methods to provide either direct produced ^99mTc or the parent ⁹⁹Mo, were studied and suggested as potential alternatives to the reactor based production of ⁹⁹Mo. A possible pathway for the charged particle production of ^99mTc and ⁹⁹Mo is irradiation of molybdenum metal with protons via the reaction ¹⁰⁰Mo(p,2n)^99mTc and ¹⁰⁰Mo(p,pn)⁹⁹Mo, respectively. The earlier published excitation functions show large differences in their maximum that result in large differences in the calculated yields. We therefore decided to study the excitation function for these proton-induced reactions. In this work the newly measured excitation functions as well as an evaluation of earlier measured data and a discussion of the observed disagreements are presented.     

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 of iodine produced from fission with a porous metal silver column in99Mo production

In Argentina, at the Ezeiza Atomic Center,¹³¹I is produced by wet distillation of natural tellurium dioxide irradiated with thermal neutrons in a pool-type reactor. In order to recover the¹³¹I present in the production process of fission⁹⁹Mo obtained by irradiation of UAL_x/Al targets (with 90% enriched uranium) a separation method was developed. Iodine isotopes can be separated from a sodium hydroxide solution containing fission products using a column filled with alternate beds of glass microspheres and porous metal silver. Tests with tracers were performed in radiochemical laboratory. Following this results, a series of tests with higher activities (3 TBq of⁹⁹Mo and 0.7 TBq of¹³¹I) were carried out in hot cells. Molybdenum passes through the silver column, while¹³¹I retention was 92–97% in tracer test and 90% in optimised hot cell tests. This result depends on several facts that are discussed. An initial separation of iodine isotopes diminishes radiation damage on ion-exchange resin used in the subsequent molybdenum purification, improving its retention and elution yield.     

A novel technique for the separation of99mTc from99Mo

A new technique for the separation of^99mTc from low specific activity⁹⁹Mo is reported. A separation based on the principle of precipitation of⁹⁹Mo as calcium molybdate has been investigated. On precipitating⁹⁹MoO ₄ ^2– from alkaline solution as calcium molybdate under controlled conditions, the^99mTcO ₄ ^– is found to remain quantitatively in the supernatant solution with little carry-over of⁹⁹Mo. This calcium molybdate (⁹⁹Mo) could be redissolved and reprecipitated at regular intervals, yielding^99mTc quantitatively in aqueous neutral solutions. Calcium molybdate precipitates containing up to 1.5 GBq of⁹⁹Mo and 130–180 mg of molybdenum were prepared and evaluated. The performance in terms of repeated^99mTc separation gave yields of 75–93% with acceptable readionuclidic and radiochemical purity.     

Feasibility study for production of 99mTc by neutron irradiation of MoO3 in a 250 kW TRIGA Mark II reactor

The subject of this paper is to explore the possibility to obtain ^99mTc from activation of ⁹⁸Mo, using the TRIGA Mark II low flux research reactor (Vienna, Austria). Irradiation of both natural and enriched in ⁹⁸Mo molybdenum oxides was compared. Aims of this work included the determination of neutron fluxes and ⁹⁸Mo(n, γ)⁹⁹Mo reaction effective cross section in the TRIGA Mark II reactor irradiation channels, calculation of ⁹⁹Mo specific activities, determination of optimal irradiation conditions for the subsequent ^99mTc separation from MoO₃ targets using concentrating technologies.     

Biosorption of 241Am by immobilized Saccharomyces cerevisiae

More than half of the world's annual production of radionuclides is used for medical purposes such as diagnostic imaging of diseases and patient therapy. Using aqueous homogeneous solution reactor technology, production quantities of medical radioisotopes ⁹⁹Mo and⁸⁹Sr, can be extracted from one reactor cycle. ⁹⁹Mo may be produced directly from UO₂SO₄ uranyl sulfate in an aqueous homogeneous solution nuclear reactor in a manner that produces high purity radionuclides, making efficient use of the reactor's uranium fuel solution. The process is relatively simple, economical, and waste free, eliminating uranium targets. The short-lived radioisotope ^99mTc is eluted from ⁹⁹Mo for diagnostic imaging. Radioisotope ⁸⁹Sr infusion is a therapeutic modality that reduces reliance on narcotic analgesia through palliation of metastatic bone pain caused by metastases of the cancer to the bone. Painful disseminated osseous metastases are common with carcinomas of the lung, prostate, and breast. Synergistic interleaving of two manufacturing processes, one producing ⁹⁹Mo and another producing ⁸⁹Sr in the same production cycle of an aqueous homogeneous solution reactor makes full and efficient use of the time for both the neutron irradiation stage and the extraction stage of each radionuclide. Interleaving the capture of ⁸⁹Sr radioisotope with production processing of ⁹⁹Mo radioisotope is achieved, since the extraction and subsequent elimination of radionuclide impurities occurs during separate parts of the reactor cycle. The process applies to either HEU or LEU nuclear fuels in an aqueous homogeneous solution reactor.     

Isotope exchange and separation factor of238U/235U for the system U(III)/U(IV) in aqueous/organic phases

A study of the isotope exchange reaction U(III)_org/U(IV)_aq in the extraction system: 7M HCl — tributyl phosphate (TBP) — toluene has been performed. For 20 s of contact the results show a separation factor²³⁵U/²³⁸U of 1.014. This large separation factor is explained by the oxidation reaction of²³⁵U(III) and²³⁸U(III).     

Dosage du molybdene et du tungstene dans les plantes par activation neutronique

A method for the determination of molybdenum and tungsten in plant material using neutron activation analysis was developed. The considered reaction are: $$\begin{gathered} {}^{98}Mo(n,\gamma )^{99} Mo \hfill \\ {}^{186}W(n,\gamma )^{187} W \hfill \\ \end{gathered} $$ The separation for tungsten and molybdenum was carried out using anion exchange separation (Dowex 1×10; 200–400 mesh). Irradiation was carried out in a swimming pool reactor at a thermal flux of about 1–2×10¹³ n·cm^?2·sec^?1 for 15 hours. The samples and standard were allowed to cool for 5–6 hours before chemical processing. the high concentration of calcium in plants (up to 40 mg/g dry material), the use of hydrofluoric acid for a good absorption and quantitative recovery of tungsten led us to dissolve the samples with Ht?H₂O₂ mixtures containing boric acid to prevent the precipitation of fluorides.     

Preparation of <Superscript>235m</Superscript>U targets for <Superscript>235</Superscript>U(n,n′)<Superscript>235m</Superscript>U cross section measurements

This paper describes the preparation of samples for an experiment to measure the cross-section for ²³⁵U(n,n′)^235mU in a fast fission spectrum of neutrons provided by a fast pulsed reactor/critical assembly. Samples of ^235mU have been prepared for the calibration of the internal conversion electron detector that is used for the ^235mU measurement. Two methods are described for the preparation of ^235mU. The first method used a U-Pu chemical separation based on anion-exchange chromatography and the second method used an alpha recoil collection method. Thin, uniform samples of ^235mU+²³⁵U were prepared for the experiment using electrodeposition.     

Cumulative yields of short-lived ruthenium isotopes in the thermal neutron induced fission of233U,235U and239Pu

Cumulative yields of short-lived ruthenium isotopes in the thermal neutron induced fission of²³⁵U,²³⁵U and²³⁹Pu have been determined using a fast radiochemical separation technique followed by gamma spectrometry. The cumulative yields of¹⁰⁷Ru and¹⁰³Ru in²³³U (n_th, f) and¹⁰⁷Ru and¹⁰⁹Ru in²³⁹Pu (n_th, f) are determined for the first time. The measured cumulative yields are converted to chain yields assuming normal charge distribution systematics for comparison with the literature data on chain yields.     

Efficacious selective separation of U(VI) over Mo(VI) using novel 2,9-diamide-1,10-phenanthroline ligands: Liquid-liquid extraction and coordination chemistry

Uranium and molybdenum are important strategic elements. The production of ⁹⁹Mo and the hydrometallurgical process of uranium ore face difficult problems of separation of uranium and molybdenum. In this study, the four phenanthroline diamide ligands were synthesized, and extraction and stripping experiments were performed under different conditions to evaluate the potential application of these ligands for separation of U(VI) over Mo(VI). With the growth of alkyl chain, the solubility of ligands could be greatly improved, and the separation effect of U(VI) over Mo(VI) gradually increased. The SF_U/Mo were around 10,000 at 4 mol/L HNO₃. Three stripping agents were tested with the stripping efficiency of Na₂CO₃ (5%) > H₂O > HNO₃ (0.01 mol/L). The stripping percentages of the three stripping agents were all close to unity, indicating that the ligands had the potential to be recycled. The chemical stoichiometry of U(VI) complexes with ligands was evaluated as 1:1 using electrospray ionization mass spectrometry, ultraviolet visible spectroscopy and single-crystal X-ray diffraction. The consistency between theoretical calculation and experimental results further explains the coordination mechanism.     

Radiochemical determination of Mo in pure tungsten

A radiochemical neutron activation technique for Mo determination in high purity tungsten, based on some specific properties of Mo and W radionuclides has been developed. Al₂O₃ powder has been used as a sorbent. An estimation of the Mo content was carried out via the selectively separated^99mTc daughter radionuclide. Limit of detection was 10 ng g^–1.     

Measurement of absolute intensity of 1001 keV gamma-ray of 234mPa

The sorption behavior of ²³⁵U fission fission products ⁹⁹Mo and ¹³²Te was studied through batch and dynamic experiments when they were dissolved in 1 to 7M HNO₃ solutions. It was found that ⁹⁹Mo is always totally adsorbed on hydrated SnO₂, while ¹³²Te is rather weakly adsorbed, therefore they can be separated from each other although ¹³²Te in the solution still remains contaminated with other radionuclides as well as ⁹⁹Mo does in the solid.     

Production of fission 131I

Summary A method of iodine separation from other radionuclides generated by ²³⁵U fission has been developed in order to explore the possibilities to obtain ¹³¹I as by-product of the ⁹⁹Mo routine production in the Ezeiza Atomic Centre. The experiments were designed to remove this element to gas phase, and the recoveries were investigated both with and without carrier addition. High volatilization percentages were achieved in the presence of iodine carrier. Some other alternatives to increase the iodine displacement to the gaseous phase, namely vacuum distillation, addition of hydrogen peroxide and use of a carrier gas, were also studied. The method developed, which employs a carrier gas stream, without carrier addition, allows the recovery of about 97% of the ¹³¹I, with high specific activity, in a simple and clean way.     

Fission yields of molybdenum in the Oklo natural reactor

The isotopic compositions of molybdenum in six uranium-rich samples from the Oklo Zone 9 natural reactor were accurately measured by thermal ionization mass spectrometry. The samples were subjected to an ion exchange separation process that removed the isobaric elements zirconium and ruthenium, with high efficiency and a low blank. Molybdenum possesses seven isotopes of which ^92,94,96Mo are unaffected by the fission process, enabling the raw data to be corrected for isotope fractionation by normalising to ⁹²Mo/⁹⁶Mo, and to use ⁹⁴Mo to correct for the primordial component in each of the fission-produced isotopes. This enables the relative fission yields of Mo to be calculated from the isotopic composition measurements, to give cumulative fission yields of 1:0.941:0.936:1.025 for ^95,97,98,100Mo, respectively. These data demonstrate that the most important nuclear process involved in reactor Zone 9 was the thermal neutron fission of ²³⁵U. The consistency of the relative cumulative fission yields of all six samples from different locations in the reactor, implies that Mo is a mobile element in the uraninite comprising Zone 9, and that a significant fraction of molybdenum was mobilized within the reactor zone and probably escaped from Zone 9, a conclusion in agreement with earlier published work.     

The determination of molybdenum and tungsten in sea and surface water

A simple method for the determination of molybdenum and tungsten in sea and surface water is presented. Molybdenum and tungsten are concentrated on activated charcoal by adsorption as the ammonium pyrrolidine dithiocarbamate complex; the optimal pH for adsorption is 1.3. Mo and W are then determined by thermal neutron activation, forming ⁹⁹Mo (T_{$\text{1}\text{2}$} = 66.7 h) and ¹³⁷W (T_{$\text{1}\text{2}$} = 23.8 h), respectively. The ^99mTc daughter of ⁹⁹Mo is measured as soon as the equilibrium between ^99mTc(T_{$\text{1}\text{2}$}= 6 h) and ⁹⁹Mo is established. The detection limits are 0.05 μg Mo l^-1 and 0.05 μg W l^-1 (or 0.001 μg W l^-1 after a simple chemical separation).     

A simple technetium generator

A systematic study on the extraction of⁹⁹Mo and its daughter^99mTc by pure organic diluents and dinonylnaphthalinesulfonic acid (DNNS) is described. The aqueous phases used are H₂SO₄, HCl, KI and their binary mixture solutions. The effect of alcohols on the distribution coefficient has been investigated. As a result of the study, a simple and rapid generator is built for the production of pure^99mTc from⁹⁹Mo.     

An alternative route for the preparation of the medical isotope 99Mo from the 238U(γ, f) and 100Mo(γ, n) reactions

The radionuclide ⁹⁹Mo, which has a half-life of 65.94 h was produced from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions using a 10 MeV electron linac at EBC, Kharghar Navi-Mumbai, India. This has been investigated since the daughter product ^99mTc is very important from a medical point of view and can be produced in a generator from the parent ⁹⁹Mo. The activity of ⁹⁹Mo was analyzed by a γ-ray spectrometric technique using a HPGe detector. From the detected γ-rays activity of 140.5 and 739.8 keV, the amount of ⁹⁹Mo produced was determined. For comparison, the amount of ⁹⁹Mo from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions was also estimated using the experimental photon flux from ¹⁹⁷Au(γ, n)¹⁹⁶Au reaction. The amount of ⁹⁹Mo from the detected γ-lines is in agreement with the estimated value for ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions. The production of ⁹⁹Mo activity from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions is a relevant and novel approach, which provides alternative routes to ^235,238U(n, f) and ⁹⁸Mo(n, γ) reactions, circumventing the need for a reactor. The viability and practicality of the ⁹⁹Mo production from the ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions alternative to ^235,238U(n, f) and ⁹⁸Mo(n, γ) reactions has been emphasize. An estimate has been also arrived based on the experimental data of present work to fulfill the requirement of DOE.