A simple method for preparation of pure 68 Ga-acetate precursor for formulation of radiopharmaceuticals: Physicochemical characteristics of the 68 Ga eluate of the SnO2 based-68 Ge/68 Ga column generator.

Gallium-68 radioisotope is an excellent source in clinical positron emission tomography application due to its ease of availability from germanium-68 (68 Ge)/gallium-68 (68 Ga) generator having a shelf life of 1 year. In this paper, a modified method for purification of the primary eluate of 68 Ge-68 Ga generator by using a small cation exchange resin (Dowex-50) column has been described. The breakthrough of 68 Ge before and after purification of 68 Ga eluate was 0.014% and 0.00027%, respectively. The average recovery yield of 68 Ga after purification was 84% ± 8.6% (SD, n = 335). The results of the physiochemical studies confirmed that the 68 Ga-acetate obtained is suitable for labeling of radiopharmaceuticals.

Preparation of [68Ga]Ga-Chloride from 68Zn solid target for the synthesis of pharmaceutical grade [68Ga]Ga-PSMA-11 and [68Ga]Ga-DOTA-TATE.

68Ga is produced from enriched zinc-68 target electrodeposited on copper base material which was irradiated with 15 MeV proton energy in 30 MeV cyclotron. A modified semi-automated separation and purification module was used to obtain pharmaceutical grade [68Ga]GaCl3 in 35 ± 5 min. The quality of [68Ga]GaCl3 produced was in accordance with Pharmeuropa 30.4. The [68Ga]GaCl3 was utilized for the formulation of multiple doses of [68Ga]Ga-PSMA-11 and [68Ga]Ga-DOTATATE. The quality of [68Ga]Ga-PSMA-11 and [68Ga]Ga-DOTATATE were also in accordance with Pharmacopeia.

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.

Development of a Ne gas target for (22)Na production by proton irradiation.

The article presents the design and development of a neon gas target for the production of (22)Na using a proton beam from the room temperature cyclotron in Variable Energy Cyclotron Centre, Kolkata. The target design is made to handle a beam power of 85 W (17 MeV, 5 µA). The design is based on simulation using the computer code FLUKA for the beam dump and CFD-CFX for target cooling. The target has been successfully used for the production of (22)Na in a 6 day long 17 MeV, 5 µA proton irradiation run.

Preparation and evaluation of SnO2-based 68Ge/68Ga generator made from 68Ge produced through (nat)Zn(α,xn) reaction.

(68)Ge was produced by (nat)Zn(α,xn)(68)Ge reaction and its production yield was 31.82 kBq/µAh (0.86 µCi/µAh) at the end of irradiation (EOI). A simple chromatographic method using a SnO2 column was employed to separate (68)Ge from the target material and the co-produced non-isotopic radioisotope impurities. (68)Ge retained in the column served as the (68)Ge/(68)Ga generator. Elution efficiency of the column was about 60%. First 2 ml of the eluate contained more than 95% of the elutable activity. Post-elution purification cum concentration was done with a small cation exchange resin column. The presence of the inactive tin ions in the (68)Ga eluate was determined by the ICP-OES technique and was found to be about 0.03 ppm. Radiochemical purity of the final (68)Ga preparation was more than 99.99% and it was found to be suitable for making complex with ethylenediamine-N,N,N',N'-tetrakis(methylene phosphonic acid) (EDTMP).

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,gamma)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 simple and rapid technique for recovery of (99m)Tc from low specific activity (n,gamma)(99)Mo 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.

Recovery of 99mTc from Na2[99Mo]MoO4 solution obtained from reactor-produced (n,gamma) 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.