The demand of radioisotopes is rising due to wide-ranging applications in industry, agriculture, medicine and in research.
Two sources of artificial radioisotopes are accelerators and reactors. The reactor offers large volume for irradiation, simultaneous
irradiation of different samples and economy of production, whereas accelerators are generally used to produce those isotopes
which can not be produced by reactor. Radioisotope production started on a significant scale in several countries with the
commissioning of research reactors starting from the late 1950s. The period from 1950 to 1970 saw construction of a large
number of research reactors with multiple facilities. After 1980, because of the decommissioning of many old ones, the number
of operating reactors has been steadily decreasing. The research reactors used for radioisotope production could be broadly
classified into swimming pool type and tank type reactors. CANDU power reactors currently produce many millions of curies
per year of 60Co for MDS Nordion’s use in industry and commerce. Studies related to production of other isotopes in power reactors have
also been performed. Indeed, while a very few reactors have come online in the past decade, many more have been retired or
may retire in coming years. After failure of MAPLE project, there has been unwillingness to built new reactors. Activism and
politics has made it so difficult to build new reactors that we are left to use only the reactors we inherited from a nuclear
era. Many design considerations and requirements for the production of isotopes in power reactors must be assessed, such as;
operator and public safety, minimum impact on station efficiency and reactor operations, shielding requirements during reactor
operation with target adjusters and removal of the target adjusters from core, transportation within the station, and finally
the processing and shipment off-site. Use of power reactors for isotope production is reviewed.
Ion-exchangers are found not only in water purification processes, the original major application, but also in analytical chemistry for the separation and isolation of elements, hydrometallurgy, inorganic chemistry and biochemistry, in food technology, and of course in many specialized fields related to the utilization of atomic energy. The use of organic ion-exchangers is limited by virtue of their limited stability under harsh conditions, whereas inorganic ion-exchangers possess important properties, which make them very useful for chemical separation and purification in intense radiation fields. The availability of short-lived radionuclides from radionuclide generators provides an inexpensive and convenient alternative to in-house radioisotope production facilities such as accelerators and cyclotrons. Due to their simplicity of operation, chromatographic based generators have been the method of choice, although generators based on solvent extraction and on volatization and sublimation have also been developed, and are routinely used. In this paper use of inorganic ion-exchangers for the development of radionuclide generators for the decade 1993–2002 has been compiled.
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.
Adsorption behavior of molybdate on acidic alumina was studied at boiling water bath temperature (~100 °C). Various parameters
affecting the adsorption of molybdenum, such as pH, amount of molybdenum, incubation period, etc., were determined. A 99mTc generator was prepared by adsorbing low specific activity 99Mo (150 mg) on 1 g alumina. Elutions were carried out with saline. Performance of the generator such as 99Mo breakthrough, aluminum contents, pH, elution profile, radiochemical purity, and labeling efficiency of kits were checked.
Lanreotide, a synthetic octapeptide analog of a native hormone somatostatin, was labeled with 131I, the most widely used therapeutic and easily available radionuclide. Radioiodination of Lanreotide was carried out by Chloramine-T
and Iodogen methods. Chloramine-T and Iodogen were used as oxidizing agents to form an electrophilic iodine species, which
then labeled the tyrosine of Lanreotide. The maximum radiolabeling yield was ~80%. Chloramine-T was found more suitable than
the Iodogen method, because nearly 25% of the initial iodine activity was lost/adsorbed on the Iodogen coating. Thin layer
and high performance liquid chromatographies were used for monitoring the reaction of 131I with Lanreotide, the stability and purity of 131I-Lanreotide.
A new formulation of a freeze-dried kit for the labeling of tetrofosmin with technetium-99m has been developed. The kit contains
lyophilized mixture of 0.320 mg tetrofosmin [6,9-bis(2-ethoxyethyl)-3,12-dioxa-6,9-diphosphatetradecane], 0.025 mg stannous
chloride dihydrate, 5 mg sodium tartrate and 5 mg sodium hydrogen carbonate. The product contains no antimicrobial preservative.
When 99mTc pertechnetate up to 6 mL saline containing 200 mCi is added to lyophilized mixture, a lipophilic, cationic 99mTc complex is formed, 99mTc-tetrofosmin. The performance of newly developed kit is compared with commercially available MYOVIEW kit for heart imaging.
The patient studies show that the images of heart obtained by 99mTc-tetrofosmin prepared by new formulation are equally good to MYOVIEW.
No-carrier-added 90Y was separated from 90Sr via colloid formation of 90Y in basic media. The mixture was passed through glass wool or membrane filter. The filtrate contained 90Sr, while 90Y was retained on glass wool/membrane filter. Yttrium-90 was extracted with 0.1 M HCl. Contamination of 90Sr was <0.0001%. More than 98% labeling yield of 90Y-EDTMP was confirmed by paper chromatography.
Sorption of Cd and Ag by a cation exchange resin has been studied at different molarities of nitric acid. The sorption capacity of Cd on a cation exchanger has been determined. A109Cd/109mAg generator is suggested, based on the sorption of Cd on AG 50W-X8 organic cation exchanger at 0.01M HNO3.109mAg is eluted with 0.2M NaCl, physiologically compatible for human use.
Authors:Tanveer Bokhari, A. Mushtaq, and Islam Khan
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.
The lyophilized MIBI kit was dissolved in 1 ml sterile saline or 250mg/ml
ascorbic acid and dispensed into 0.2 ml fractions, which were stored at -20 °C for 12 days. The solution was prepared by using
two different protection methods. In the first method evacuated vials were used for storage of fractionated solution while
in the second method an antioxidant agent, ascorbic acid was employed. The radiochemical impurity of 99mTc-MIBI in the unprotected fractions rises with time. Exclusion of air as well addition of ascorbic acid in fractionated solutions
gave very good results. The labeling efficiency and biodistribution of fractionated solutions was the same as the lyophilized
kit even after 12 days.