Sujata Saha Das

Synthesis and evaluation of 99mTc-moxifloxacin, a potential infection specific imaging agent

To synthesize and evaluate a (99m)Tc labeled fluroquinolone, moxifloxacin as a potential bacteria specific infection imaging agent. A radiolabeling formulation including moxifloxacin, [Moxicip(TM) injection, Cipla] (4mg), sodium pertechnetate and stannous chloride (5microg) gave the best radiolabeling efficiency and moderately stable labeled (99m)Tc moxifloxacin. Quality control analysis was performed by ITLC. Rats and rabbit with infectious intramuscular lesions induced in either thigh with E. Colli were used for studying biodistribution and scintigraphic imaging of the labeled product. Imaging of an infected thigh of a rabbit was performed with a gamma-camera at various intervals. A good radiolabeling efficiency (90-95%) was obtained within 5min. No purification of the labeled product was done. Labeled product retained its radiochemical purity upto 85% even at 3h. Scintigraphy showed uptake in infectious lesions at 30min after injection, which remains constant upto 3h study. Abscess-to-muscle ratios were 1.60, 1.62, 1.74 and 1.75 at 30min, 1, 2 and 3h, respectively. Thus, (99m)Tc moxifloxacin, a new potential radiopharmaceutical has been developed for infection imaging agent.

Recovery of 99mTc from Na2[99Mo]MoO4 solution obtained from reactor-produced (n,γ) 99Mo using a tiny Dowex-1 column in tandem with a small alumina column☆

A simple separation technique of (99m)Tc from Na(2)[(99)Mo]MoO(4) in sodium hydroxide solution obtained from the (98)Mo(n,gamma)(99)Mo reaction is described. Low to medium specific activity (99)Mo-molybdate solution of 7.4-18.5GBq (200-500mCi) in sodium hydroxide was passed through a tiny Dowex-1 column (25mg) to separate the (99m)Tc from the (99)Mo; subsequently the (99m)Tc was eluted from the Dowex 1 column with tetrabutylammonium bromide (TBAB) solution (1mg/5ml methylene chloride). The TBAB solution was passed through a small alumina column (1.5g) where the (99m)Tc is retained and separated from TBAB and CH(2)Cl(2). Technetium-99m from the alumina column was finally eluted with 5ml saline leaving any traces of (99)Mo on the alumina column. The separation yield was about 90% (n=10). The method has applicability for decontamination of (99g)Tc from spent (99)Mo waste solution and recovery of (99g)Tc for research use. The procedure should also be equally applicable for recovery of (188)ReO(4) from (188)WO(4) in a radioisotope laboratory.

A simple and rapid technique for recovery of 99mTc from low specific activity (n,γ)99Mo based on solvent extraction and column chromatography

A simple and inexpensive method of separation of (99m)Tc from (99)Mo produced by neutron activation of (98)Mo via the (98)Mo(n,gamma)(99)Mo nuclear reaction is described. The recovery of (99m)Tc was performed by solvent extraction technique followed by column (active alumina) chromatography. The overall radiochemical yield for the complete separation of (99m)Tc was 85-95% (n=10). The separated Na[(99m)Tc]TcO(4) was of high radionuclidic, radiochemical, and chemical purities. The method can be adopted for routine use of (99m)Tc in hospital radio-pharmacies utilizing low-medium specific activity (n,gamma)(99)Mo produced in a research reactor.

A computerized compact module for separation of 99mTc-radionuclide from molybdenum

An automated closed cycle module for the separation and recovery of various isotopes, radioactive or non-radioactive, using solvent extraction and column chromatography techniques, and in particular, for separation and recovery of (99m)Tc from low-medium specific activity (99)Mo, is described. The module may also be applicable for separation of (99m)Tc produced in a cyclotron. The module is safe and reliable to avoid human interference and hazards posed by handling of radioactive and hazardous chemicals. The entire system of automation includes a user-friendly PC based graphical user interface (GUI) that actually supervises the process via an embedded system based electronic controller.

Recovery of 131I from alkaline solution of n-irradiated tellurium target using a tiny Dowex-1 column

A simple and inexpensive ion-exchange chromatography method for the separation of medically useful no-carrier-added (nca) iodine radionuclides from bulk amounts of irradiated tellurium dioxide (TeO(2)) target was developed and tested using (131)I. The radiochemical separation was performed using a very small Dowex-1x8 ion-exchange column. The overall radiochemical yield for the complete separation of (131)I was 92+/-1.8 (standard deviation) % (n=8). The separated nca (131)I was of high, approximately 99%, radionuclidic and radiochemical purity and did not contain detectable amounts of the target material. This method may be adopted for the radiochemical separation of other different iodine radionuclides produced from tellurium matrices through cyclotron as well as reactor irradiation.

A simple and rapid technique for radiochemical separation of iodine radionuclides from irradiated tellurium using an activated charcoal column.

A simple and inexpensive method for the separation of medically useful no-carrier-added (nca) iodine radionuclides from bulk amounts of irradiated tellurium dioxide (TeO(2)) target was developed. The beta(-) emitting (131)I radionuclide, produced by the decay of (131)Te through the (nat)Te(n, gamma)(131)Te nuclear reaction, was used for standardization of the radiochemical separation procedure. The radiochemical separation was performed by precipitation followed by column (activated charcoal) chromatography. Quantitative post-irradiation recovery of the TeO(2) target material (98-99%), in a form suitable for reuse in future irradiations, was achieved. The overall radiochemical yield for the complete separation of (131)I was 75-85% (n=8). The separated nca (131)I was of high, approximately 99%, radionuclidic and radiochemical purities and did not contain detectable amounts of the target material. This method can be adopted for the radiochemical separation of other different iodine radionuclides produced from tellurium matrices through cyclotron as well as reactor irradiation.

Production of 61Cu using natural cobalt target and its separation using ascorbic acid and common anion exchange resin.

(61)Cu was produced by (nat)Co(α, xn)(61)Cu reaction. (61)Cu production yield was 89.5 MBq/μAh (2.42 mCi/μAh) at the end of irradiation (EOI). A simple radiochemical separation method using anion exchange resin and ascorbic acid has been employed to separate the product radionuclide from inactive target material and co-produced non-isotopic impurities. The radiochemical separation yield was about 90%. Radiochemical purity of (61)Cu was >99% 1 h after EOI. Final product was suitable for making complex with N(2)S(2) type of ligands.

A simple and rapid technique for recovery of 99mTc from low specific activity (n,γ)99Mo based on solid–liquid extraction and column chromatography methodologies

A simple and inexpensive method has been developed for the separation of (99m)Tc from (99)Mo produced from the neutron activation of (98)Mo by (98)Mo(n,gamma)(99)Mo nuclear reaction. The recovery of (99m)Tc was performed by solid-liquid extraction based on alumina column chromatography. The overall radiochemical yield for the complete separation of (99m)Tc was 85-97% (n=5). The separated Na[(99m)Tc]TcO(4) was of high radionuclidic, radiochemical and chemical purities. The method can be adopted for routine processing and use of (99m)Tc in radiopharmacy operations.

A compact solvent extraction based 99Mo/99 mTc generator for hospital radiopharmacy

A compact and portable 99Mo-99 mTc generator based on solvent-extraction, mimic to the conventional 99Mo-99 mTc alumina column generator is much-needed commodity for use in hospital radiopharmacy setup. The present study includes the development of a portable, simple and low cost 99Mo/99 mTc-generator based on MEK solvent extraction technique to obtain a very high concentration of no-carrier added (nca) 99 mTc solution, where low specific activity 99Mo source is obtained through 98Mo(n, γ)99Mo reaction in a research reactor. The unit is intended for operation under the conditions of medical radiological laboratories. Technical trials showed that the mean time of preparation of sodium [99mTc] pertechnetate radiopharmaceutical did not exceed 15 min. The quality and yield of 99 mTc-pertechnetate is upto the mark for formulation of radiopharmaceuticals.

Radioactive ion beams of 111In using ECR plasma sputtering method

Radioactive ion beams of 111In (indium-111, half-life 2.8 days) have been produced using the plasma sputtering method in an electron cyclotron resonance (ECR) ion source at the Variable Energy Cyclotron Centre RIB facility. Indium isotopes were first produced by bombarding a natural silver target with a 32 MeV, 40 μA alpha particle beam from the K-130 cyclotron. After radio-chemical separation, about 25 mCi In-chloride was deposited on an aluminum electrode and inserted in the plasma chamber of the ECR. Indium ions produced by ion induced sputtering in the plasma were extracted from the ion source, isotopically separated, and a pure 111In beam was measured at the focal plane of the separator. The measured 111In beam intensity was 2.67 × 105 particles/s for a beam energy of 5 keV.