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.     

The study of conditions for the preparation and application of99Mo−99mTc generators starting from irradiated molybdenum metal

⁹⁹Mo−^99mTc generators were prepared starting from irradiated molybdenum metal instead of MoO₃ in order to use reactor irradiation space more economically. The adsorption of molybdenum as sodium molybdate on different kinds of alumina was investigated. The effect of the pH of the column and the aqueous phase concentration of molybdenum were studied and related to the elution yield of^99mTc. A study of the radiation damage effect indicated that generators having a high elution efficiency of 90% could be prepared in the 100–600 mCi range. The losses of⁹⁹Mo were minimized to 10⁻⁵-10⁻⁴% and those of alumina to 2–5 μg/ml eluate.     

Rapid and sensitive determination of pertechnetate in99Mo/99mTc generator eluates by reversed-phase high-performance liquid chromatography

Reversed-phase high-performance liquid chromatography with u.v. detection was applied for rapid and sensitive determination of pertechnetate in⁹⁹Mo/^99mTc generator eluates, using a mixture solvent of acetonitrile and 0.04M aqueous acetate buffer (1/1) containing a few volume percentage of 0.5 M tetra-n-butylammonium hydroxide as the mobile phase. Employing a -bondapak C₁₃ column, the TcO ₄ ^– species was separated, monitored with absorbance at 254 nm, and observed at the retention time of 3.5 min. The detection limit was found to be 5.2·10^–10 g of Tc for each injection. Total Tc contents in the^99mTc eluates from clinically-used⁹⁹Mo/^99mTc generator were analyzed by this technique. The^99mTc (⁹⁹Tc) species was separated from the contaminant⁹⁹Mo. This method was found to be useful for the purification of^99mTc (⁹⁹Tc) as well as the determination of total Tc content.     

Statistical analysis of biodistribution of99mTc-colloid preparations

The results of radiochemical purity measurements for^99mTc–S colloid, 90.7±1.4, and^99mTc–Sn colloid, 98.9±1.2 obtained by ITLC (SG) with 80% methanol are given. The range of biodistribution normal values for^99mTc colloid preparations for animal organs is determined. The results of^99mTc–S and^99mTc–Sn colloid distribution in liver are 95.4±6.1% and 100,0±5.4%, respectively.     

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.     

Separation of 99mTc from 99Mo by Using TOPO — Kerosene Supported Liquid Membrane

Transport of ^99mTcO₄ ^– ions across TOPO-kerosene based supported liquid membrane was investigated at different concentrations of phosphoric acid as a feed solution and different concentrations of TOPO in the membrane, where 0.9% NaCl aqueous solution was used as a stripping solution. The flux of TcO₄ ^– ions across this liquid membrane varied with the concentration of both H₃PO₄ and TOPO. The best permeability coefficient was obtained at concentrations, [H₃PO₄] = 3 mole·l^–1 and [TOPO] = 0.5 mole·l^–1 (P = 2.08·10^–9 m²·s^–1). The results were utilized for the separation of ^99mTc from ⁹⁹Mo, where a selective and effective separation was obtained since no ⁹⁹Mo transport across this liquid membrane was noticed while a high rate of ^99mTc transport took place.     

A novel electrochemical <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator

A novel electrochemical process to avail clinical grade ^99mTc from (n,γ)⁹⁹Mo has been demonstrated. The electrochemical parameters were optimized to maximize the ^99mTc yield with minimal ⁹⁹Mo contamination. ⁹⁹Mo/^99mTc generators containing up to 29.6 GBq (800 mCi) ⁹⁹Mo were developed and their performance were extensively evaluated for 10 days without changing the operating conditions. Very high radioactive concentration of ^99mTcO₄ ⁻ of acceptable quality, commensurate with hospital radiopharmacy requirements could be availed from the system with >90% yield. The compatibility of the product for the formulation of ^99mTc labeled radiopharmaceuticals such as ^99mTc-DMSA and ^99mTc-EC was found to be satisfactory in terms of high labeling yields. The proposed route represents an important step for enhancing the scope of accessing clinical grade ^99mTc from low specific activity (n, γ)⁹⁹Mo.     

Technetium-99m generators based on neutron irradiated 12-molybdocerate as column matrix

Neutron irradiated 12-molybdocerate(IV) is evaluated as a column matrix for use in preparation of small chromatographic column type^99mTc generators. Greater than 87% of the generated^99mTc activity is immediately and reproducibly eluted with passing 10 cm³ of saline or chromated saline solutions through 3.0 g of molybdocerate column bed at flow rates of about 1.0 cm³/min. The elution performance of^99mTc does not drastically change with increasing the drying temperature and/or the particles sixe of the column matrix in the range investigated. Saline eluent containing 5·10^–5M CrO ₄ ^2– increases the radiochemical purity of the eluate to >98% TcO ₄ ^– . Radionuclidic and chemical purities of the eluates satisfy the specifications for use in nuclear medicine.     

Technetium-99m generator based on 12-molybdocerate —99Mo precipitate as column matrix

The preparation of insoluble 12-molybdocerate(IV) from⁹⁹Mo of low specific activity, produced by thermal neutron irradiation of MoO₃, is described. Samples of the material are dried at 50, 100 and 200°C and used as column matrices from which the generated^99mTc activity is periodically eluted with saline solution or saline solution containing 5·10^–5M K₂CrO₄ as an oxidant. The elution yields of^99mTc are high and reproducible (95–81%) with radionuclidic purity 99.98%. Both chemical and radiochemical purity (as TcO ₄ ^– ) of the eluates decrease with increasing drying temperature of the column matrix. Using chromated saline solution as eluent improves the radiochemical purity of the^99mTc eluate.     

Practicality of Tetragonal Nano-Zirconia as a Prospective Sorbent in the Preparation of 99Mo/99mTc Generator for Biomedical Applications

The feasibility of using tetragonal nano-zirconia (t-ZrO₂) as an effective sorbent for developing a ⁹⁹Mo/^99mTc chromatographic generator was demonstrated. The structural characteristics of the sorbent matrix were investigated by different analytical techniques such as XRD, BET surface area analysis, FT-IR, TEM etc. The material synthesized was nanocrystalline, in tetragonal phase with an average particle size of ~7 nm and a large surface area of 340 m² g^?1. The equilibrium sorption capacity of t-ZrO₂ is >250 mg Mo g^?1. The present study indicates that ⁹⁹Mo is both strongly and selectively retained by t-ZrO₂ at acidic pH and ^99mTc could be readily eluted from it, using 0.9% NaCl solution. A 9.25 GBq (250 mCi) t-ZrO₂ based chromatographic ⁹⁹Mo/^99mTc generator was developed and its performance was repeatedly evaluated for 10 days. ^99mTc could be eluted with >85% yield having acceptable radionuclidic, radiochemical and chemical purity for clinical applications. The compatibility of the product in the preparation of ^99mTc labeled formulations such as ^99mTc-EC and ^99mTc-DMSA was evaluated and found to be satisfactory.     

Determination of alpha and pure beta emitting impurities in99Mo/99mTc generator eluates

The determination of α- and pure β-emitting impurities in⁹⁹Mo/^99mTc generator eluates before application is discussed. It is shown that α-emitters in ≥10 mCi eluates can be determined in a proportional counter or an α-spectrometer within 30 min with detection limits lower than the European Pharmacopoeia limit (10⁻⁷% of the main activity).⁸⁹Sr and⁹⁰Sr are determined after chemical separation, consisting of ion exchange to separate the main activity (TcO ₄ ⁻ ), shaking with charcoal to separate colloidal particles containing Mo and precipitation of Sr as oxalate or sulfate. The precipitate is measured in a low-level counter. For eluates with an activity of ≥10 mCi the detection limits are one order of magnitude lower than the activity limit given by the European Pharmacopoeia for⁹⁰Sr (6×10⁻⁶% of the main activity). The time needed for chemical separation and activity measurement is about 30 min.     

Extraction of99mTc into MEK from large quantity of molybdate retained on alumina column

A column-extraction generator based on the adsorption of alkaline sodium molybdate on an acid alumina column is described. Generators containing 1 to 3.2 g of molybdenum /⁹⁹Mo/ absorbed on 12–15 g of acid alumina, with activities ranging from 0.52 to 5.2 GBq /14 to 140 mCi/, have been prepared and evaluated. The results indicate that^99mTc can be quantitatively eluted from such columns with about 10 ml of MEK. The yields were found to be in the range of 70 to 90% over 3–6 days.     

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.     

A99Mo-99mTc generator based on the use of zirconium molybdophosphate-99Mo gel

A modified⁹⁹Mo–^99mTc gel generator is described. The present generator uses an insoluble zirconium molybdophosphate (ZrMP) gel tagged with⁹⁹Mo. Molybdenum-99 is chemically combined in the gel structure and cannot be eluted from the matrix. The presence of phosphate increases the chemical stability of the gel and decreases the molybdenum breakthrough. The prepared gel is sufficiently porous to permit ready diffusion of^99mTcO ₄ ^– which can be cluted with saline in yields of up to 90%. The gel was found to contain 25.1% Mo, 21.9% Zr, and 0.7% P in a molar ratio of 1.09:1.0:0.09, respectively. The high molybdenum content of the gel allows the use of cheap, non-polluting (n, )⁹⁹Mo. The eluted^99mTc was of high purity and can be used for medical and pharmaceutical applications.     

Separation of99mTc from neutron-irradiated MoO3 by precipitation as CaMoO4

A method is discussed for separating^99mTc from⁹⁹Mo by dissolution and reprecipitation of CaMoO₄ containing⁹⁹Mo-^99mTc.^99mTc can be obtained successfully with a yield of about 80% and with a radionuclidic purity as high as over 99.7%. On the other hand, by agitating Ca⁹⁹ MoO₄ solid with 0.9% NaCl aqueous solution,^99mTc can directly be obtained with a yield of about 30% without any impurities.     

Determination of technetium-99 in environmental samples by solvent extraction at controlled valence

Distribution coefficients of technetium and ruthenium are determined under different conditions with CCl₄, cyclohexanone, and 5% tri-isooctylamine (TIOA)/xylene. A method for analyzing⁹⁹Tc in environmental samples has been developed by solvent extraction in which the valences of technetium and ruthenium are controlled with H₂O₂ and NaClO. Technetium and ruthenium which are oxidized to TcO ₄ ^– and RuO ₄ ^– by NaClO are separated by extraction with CCl₄ at pH 4. The RuO ₄ ^– is reduced to low valence and technetium is kept in the TcO ₄ ^– state with H₂O₂. Technetium, ruthenium, and other nuclides are subsequently separated by solvent extraction with cyclohexanone and 5% TIOA/xylene. The decontamination of the procedure is 1.35·10⁵ for¹⁰³Ru and 1.66·10⁵ for^110mAg. The chemical yield of technetium-99 is 55%.On leave from Institute of Atomic Energy, Beijing, China     

Preparation of99Mo,132Te isotopes and99mTc,132I generators

Adsorption and desorption of⁹⁵Zr−⁹⁵Nb,⁹⁹Mo,¹⁰³Ru,¹³²Te and²³⁹Np in a HCl-alumina system were studied in order to purify⁹⁹Mo and¹³²Te obtained by the cation-exchange separation of fission products and to prepare highly pure^99mTc and¹³²I generators.⁹⁹Mo and¹³²Te, of which radionuclidic purity was over 99.99% and 99.999%, respectively, could be obtained by passing the cation-exchange separated Mo and Te fractions through alumina columns, by washing with HCl and finally by eluting⁹⁹Mo with 1M NH₄OH and¹³²Te with 3M NaOH. In order to raise the recovery of⁹⁹Mo and¹³²Te from the alumina columns, they should be eluted as quickly as possible after the adsorption. The direct use of the alumina column containing⁹⁹Mo or¹³²Te as the generator allowed milking of^99mTc or¹³²I, of which radionuclidic purity was over 99.999%. Milking yields of^99mTc with 0.1M HCl and¹³²I with 0.01M NH₄OH were 77% and 90%, respectively. The latter value was much higher than that in usual performance of the generator.     

Hydrated titanium dioxide as an adsorbent for99Mo−99mTc generator

The adsorption behaviour of⁹⁹Mo in the form of molybdate and of^99mTc in the form of pertechnetate on hydrated titanium dioxide was investigated at different molarities of hydrochloric acid. The adsorption capacity of molybdate on hydrated TiO₂ is higher than on Al₂O₃. A^99mTc-generator is suggested. This generator is based on the adsorption of (⁹⁹Mo) molybdate on hydrated TiO₂, at acidities of 0.05–0.1M. HCl.^99mTc is eluted with 0.9% NaCl. Radionuclidic, radiochemical and chemical purities of the eluates were checked. This generator seems to have a great potential as compared to the traditional alumina generators.     

Radiochenical determination of technetium-99 in LLW by chelation with sodium diethyl dithiocarbamate (NaDDC) and extraction with chloroform

A radiochemical method has been developed for the determination of⁹⁹Tc in low-level radioactive, waste from nuclear facilities, using^99mTc as an internal tracer. Radioactive contaminants were removed by carrier hydroxide precipitation and chelating extraction with NaDDC/CHCl₃ system at pH 4. The final technetium was chelated with NaDDC in 3N HCl solution and extracted selectively into chloroform. The average of radiochemical recovery for various types of LLW sample is about 90%. The decontamination factors for most radioactive nuclides are higher than 10⁵. The detection limit for⁹⁹Tc in a sample of about 10 g is 0.17 pCi/g (6.5 Bq·kg^–1) for a 100-minute count.     

Developing a Chromatographic 99mTc Generator Based on Mesoporous Alumina for Industrial Radiotracer Applications: A Potential New Generation Sorbent for Using Low-Specific-Activity 99Mo

The commercial low-pressure column chromatographic ⁹⁹Mo/^99mTc generator represents a reliable source of onsite, ready-to-use ^99mTc for industrial applications. These generators use fission-produced ⁹⁹Mo of high specific activity, posing serious production challenges and raising proliferation concerns. Therefore, many concepts are aimed at using low-specific-activity (LSA) ⁹⁹Mo. Nonetheless, the main roadblock is the low sorption capacity of the used alumina (Al₂O₃). This study investigates the feasibility of using commercial alumina incorporated with LSA ⁹⁹Mo to develop a useful ⁹⁹Mo/^99mTc generator for industrial radiotracer applications. First, the adsorption profiles of some commercial alumina sorbents for LSA ⁹⁹Mo were tested under different experimental conditions. Then, the potential materials to develop a ⁹⁹Mo/^99mTc generator were selected and evaluated regarding elution yield of ^99mTc and purity. Among the sorbents investigated in this study, mesoporous alumina (SA-517747) presented a unique sorption-elution profile. It demonstrated a high equilibrium and dynamic sorption capacity of 148 ± 8 and 108 ± 6 mg Mo/g. Furthermore, ^99mTc was eluted with high yield and adequate chemical, radiochemical, and radionuclidic purity. Therefore, this approach provides an efficient and cost-effective way to supply onsite ^99mTc for radiotracer applications independent of fission-produced ⁹⁹Mo technology.