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.     

Radiochemical separation of99Mo from natural uranium irradiated with thermal neutrons

⁹⁹Mo was separated from uranium and insoluble fission product hydroxides. More than 98% of⁹⁹Mo radioactivity was extracted with bis (2-ethylhexyl)phosphoric acid. The organic phase was washed and⁹⁹Mo was back-extracted from the organic phase with NH₄OH solution. The percent recovery from the organic phase was 91% and the purity of⁹⁹Mo was more than 99%. Pure^99mTc was also extracted from the organic phase with a saline solution. Reversed-phase partition chromatography was used for the purification of⁹⁹Mo from¹³¹I and other fission products (10% HDEHP on kieselguhr bed).¹³¹I and other isotopes were quantitatively eluted with 0.1M H₂SO₄,⁹⁹Mo was eluted using a mixture of 0.5 M HCl and 30% H₂O₂.     

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.     

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.     

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.     

Adsorption of fission products on a metal surface in nitric acid solutions: Radiochemical study using a multitracer

Adsorption behavior of fission products in nitric acid solution on various alloys and metals was studied by using a multitracer produced by neutron irradiation of UO₂. The adsorption behavior of the fission products ⁹⁹Mo, ¹³¹I, ¹³²Te, ¹⁴⁰La, and ¹⁴³Ce, and ²³⁹Np was simultaneously studied. Some chemical decontamination tests were also examined. Clear adsorption of ⁹⁹Mo, ¹³¹I, and ¹³²Te was observed, whereas adsorption of ¹⁴⁰La, ¹⁴³Ce, and ²³⁹Np was not. The adsorption characteristics were discussed by considering anion-exchange reaction and surface complexation.     

Determination of calibration curves for <Superscript>131</Superscript>I in thyroid tumour metabolic radiotherapy and other radionuclides used in SPECT imaging

Large columns containing aluminum oxide (Al₂O₃) or gel (e.g. zirconium molybdate) are needed to prepare ⁹⁸Mo(n,γ)⁹⁹Mo→^99mTc column chromatographic generators that results in large elution volumes containing relatively high ⁹⁹Mo impurity and low concentrations of ^99mTc. The decrease in radioactive concentration or specific volume concentration of ^99mTc places a limitation on some pharmaceutical kits (DTPA, MIBI, ECD, etc.) or clinical procedures. We report on the post elution concentration of ^99mTc using in house prepared lead cation-exchange and alumina columns. Using these columns high bolus volumes (10–60 mL 0.02M sodium sulfate) of ^99mTc can conveniently be concentrated in 1 mL of physiological saline. This approach also works very effectively to prepare high specific volume solutions of ^99mTc-pertechnetate from a fission based ⁹⁹Mo/^99mTc generator in the second week of its normal working life.     

<Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators based on aluminum molybdate gel matrix prepared by nano method

A procedure for preparation of ⁹⁹Mo/^99mTc radioisotope generator based on low specific activity neutron activated ⁹⁹Mo was developed. Aluminum molybdate(VI)-⁹⁹Mo of high Mo(VI) content (~?364 mg/g Al⁹⁹Mo) was prepared by mixing low specific activity molybdate(VI)-⁹⁹Mo and aluminum mixture solution with isoamyl alcohol. Al⁹⁹Mo gel matrix was precipitated when the pH of the mixture solution was raised to ~?5 by addition of NaOH to the mixture. Radiometric measurements indicate the strong fixation of Molybdate(VI)-⁹⁹Mo species in the form of the sparingly insoluble Al⁹⁹Mo gel matrix. The prepared AlMo gel matrix was physiochemically characterized. Al⁹⁹Mo gel matrix was used as a base material for preparation of ⁹⁹Mo/^99mTc generator. The ^99mTc eluted from ⁹⁹Mo/^99mTc radioisotope generator was found to have relatively high elution yield (84?±?2.3%), radionuclidic (≥?99.99%), radiochemical (98.1?±?0.9%) and chemical purity.     

Preparation of a sterile closed system99mTc generator based on zirconium molybdate

A procedure for preparation of a sterile closed system generator for^99mTc based on conversion to zirconium molybdate of⁹⁹Mo produced by neutron activation is reported. The generator is sterilized by autoclaving.^99mTc is eluted using 0.9% NaCl with high yield and purity in successive elutions.     

99mTc generator with hydrated MnO2 as adsorbent of 99Mo

An ^99mTc generator with MnO₂ as adsorbent of ⁹⁹Mo was investigated. Through batch experiments the retention of ⁹⁹Mo onto MnO₂was studied as a function of the shaking time and the pH value of the ⁹⁹Mo solution. It was found that ⁹⁹Mo showed a retention of 100% onto MnO₂ at the pH range from 3 to 11, and the equilibrium distribution was reached in less than 10 minutes. In column experiments the retention of ⁹⁹Mo onto MnO₂ was also high: 99.72%. In this case, the pH of the utilized ⁹⁹Mo solution was 5. The ^99mTc could be eluted from the MnO₂-⁹⁹Mo column by using either distilled water with a pH of 5 or an aqueous solution of 0.9% NaCl. With the saline solution, the ^99mTc elution yields were higher than 80%, and only one aliquot of 5 ml was needed to get these yields. The best results were obtained when the column was packed with 1 g of MnO₂. The water and the saline solution were passed through the column at a rate of 1.25 ml/min.     

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.     

A modified sorbent for99mTc generator

A modified sorbent for^99mTe generators of higher activities has been developed. The sorbent consists of two layers. The layer in which (F.P.)⁹⁹Mo is adsorbed contains alumina and silica gel mixture in the weight ratio 40∶60%. The underlaying layer contains 0.5% g of pure alumina. The performances of the columns filled with this sorbent are compared to these containing pure alumina with respect to the total elution efficiency of^99mTc and the elution efficiency ratio of subsequent elutions. Radiochemical and radionuclidic purities (⁹⁹Mo breakthrough) of eluates from both kinds of columns have also been determined and compared.     

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.     

Chemical and radiochemical evaluation of the purity of99mTc extracted by MEK

Solvent extraction separation of^99mTc from⁹⁹Mo using methyl ethyl ketone(MEK) has been found to be an effective method of obtaining^99mTc of medicinal purity from low specific activity⁹⁹Mo. The authors have investigated the effect of alkali and molybdenum concentration on the extraction of⁹⁹Mo and^99mTc into methyl ethyl ketone. The possibility of methyl ethyl ketone forming enol and condensation products and its effect on the final extraction efficiency and purity of^99mTc has been studied. Sodium molybdate has been found to have a good salting out effect on^99mTc pertechnetate and hence^99mTc extraction can be better accomplished from low specific activity⁹⁹Mo solutions. The ketone seems to form traces of condensation products in the extraction procedure. These have been found to be coextracted with^99mTc into MEK but did not affect the extractability of^99mTc. It was observed that neutral alumina column removes these condensation products from MEK containing^99mTc. Alternately these could be filtered off by acidification of the final aqueous^99mTc solution. The studies indicate that under optimum experimental conditions methyl ethyl ketone separates^99mTc from⁹⁹Mo with high efficiency and yields^99mTc of high purity suitable for use in nuclear medicine in the form of various labelled compounds.     

Production of99Mo and its application in nuclear medicine

Molybdenum-99 is produced in large quantities as the parent radioisotopes of^99mTc, which has been used recently in nuclear medicine. The neutron capture reaction on molybdenum and the nuclear fission of uranium are used for the large scale production of⁹⁹Mo. The products by these methods are used properly according to the objects of diagnosis. In this paper, the production of⁹⁹Mo is reviewed and the development of the production in JAERI is outlined.     

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.     

Preparation and characterization of an extraction chromatography column for technetium separation based on Aliquat-336 and silica gel support

An extraction chromatographic material based on Aliquat-336 anchored on hydrophobized silica gel support was prepared as an ion exchanger. The prepared material appeared to be suitable for the separation of ⁹⁹Tc from environmental matrices in column application. The properties of the material, sorption characteristic and distribution coefficient of ^99mTcO₄ ^-in various media were studied. The prepared sorbent was conditioned by washing with nitric acid. The solution containing ^99mTcO₄ ^- in 0.1M HNO₃ was passed through the column. Tc was eluted from the column by 8M HNO₃. The flow rate was 0.4 ml/min. The chemical yield of technetium during the separation process was determined using ^99mTc tracer and gamma measurement. The sorption recovery of Tc from the prepared sorbent with 0.1M HNO₃ solution was more than 98%, and the desorption recovery from the column using 8M HNO₃ varied between 92-96%. It was found that the prepared sorbent is suitable for the separation of technetium from environmental matrices and radioactive wastes. This revised version was published online in July 2006 with corrections to the Cover Date.     

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 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.     

Precision measurements of the half-lives of60mCo,79mSe,104mRh,149Nd,176mLu,177Lu and198Au

Carrier-free^99mTcO₄ ^– eluated from clinical⁹⁹Mo/^99mTc generators was examined to determine its specific activity. The observed specific activity was found to be always lower than the calculated value based on the⁹⁹Mo–^99mTc–⁹⁹Tc decay scheme data. These results could be explained in terms of existence of excess Tc loaded onto generator column.