Production of pharmaceutical grade [201Tl]Thallous chloride using 30 MeV cyclotron.

Multiple patient doses of [201Tl]TlCl has been produced using electrodeposited enriched 203Tl in 30 MeV cyclotron (Cyclone-30) with 28 MeV proton energy at 50 µA beam current for 8 h. Ion Exchange Column Chromatography (IECC) and liquid-liquid extraction has been employed for semi-automated radiochemical separation and purification of produced [201Tl]TlCl. The produced [201Tl]TlCl was used in coronary artery disease (CAD) patients.

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.

Clinical doses production of pharmaceutical grade Sodium[18F]Fluoride using modified integrated fluidic processor in a SYNTHERA® module.

A fully automated large-scale production of sodium [18F]fluoride ([18F]NaF) using SYNTHERA module with a modification in integrated fluidic processor (IFP) is reported. This modified IFP module is used to prepare [18F]NaF with more than 98% non-decay corrected radiochemical yield (RCY) within 5 min with specifications in accordance with United State Pharmacopeia (USP) monograph. The graphical user interface (GUI) is designed to perform the synthesis steps either manually or automatically and give information to the operator during the course of production. The desired clinical results add support to indigenously produced [18F]NaF as a pharmaceutical grade diagnostic radiopharmaceutical.

Clinical Efficacy of Sodium [99mTc] Pertechnetate from Low Specific Activity 99Mo/99mTc Autosolex Generator in Hospital Radiopharmacy Centre.

BACKGROUND: Few nuclear reactors in the world producing high specific activity (HSA) 99Mo using enriched 235U (HEU), are aging and are planned for shut down in the near future. Further, HEU will not be freely available, due to safeguards, and the technology for 99Mo from low-enriched 235U (LEU) is not yet widely accepted since 239Pu contamination in the product is an issue. Production of 99mTc from low specific activity (LSA) 99Mo obtained from 98Mo(n,)99Mo reaction in research reactor and 100Mo(,n)99Mo reaction in accelerator or directly from 100Mo(p,2n)99mTc nuclear reaction in cyclotron, has been explored [1]. The methyl ethyl ketone (MEK) based solvent extraction technique is n well known method for the separation of 99mTc from low specific activity 99Mo. The 99Mo/99mTc autosolex generator [2], a computer controlled automated module, utilizes the conventional MEK solvent extraction method for extraction of 99mTc. Herein, we have validated the usage of autosolex for preparation of pharmacopoeia grade 99mTcO4- from 7.40-27.5 GBq of LSA 99Mo-SodiumMolybdate (99MoO42-) solution and validated the quality of the 99mTcO4- by preparing wide range of 99mTc-radiopharmaceuticals (99mTc-RP). MATERIALS AND METHODS: The 99mTcO4- was extracted from the autosolex as described in [2] starting from 7.40-27.5 GBq of LSA 99MoO42- and subjected to the required physico-chemical and biological quality control (QC) tests. The eluted 99mTcO4- labeled various fourth generation 99mTc radiopharmaceuticals cold kits (99mTc-cold kits) apart from regular 99mTc-cold kits in our centre. Various 99mTc-RP extracted 99mTcO4- using standard procedures [3] were prepared and subjected to required QC as Indian Pharmacopeia monograph [4] and used in scintigraphic imaging in patients. The radiation exposure dose to the operator were compared between autosolex and manual MEK based solvent extraction generator. RESULTS: The extracted 99mTcO4- from autosolex is a clear and colorless solution with pH between 5.0-6.5. The elemental molybdenum (Mo) and aluminum (Al) content <10µg/mL, MEK levels <0.1%, 99Mo breakthrough <0.030% and radiochemical purity (RCP) >98%. All the extracted 99mTcO4- batches complies sterility test, endotoxin limit (EL) <5EU/mL. The RCP of all the labeled 99mTc-RP >95%. The autosolex delivers much less radiation dose to the operator than the convention manually handled MEK based solvent extraction generator. CONCLUSIONS: Autosolex Generator was successfully used to obtain pharmaceutical grade 99mTcO4- from LSA 99MoO42- and generator is safe in radiological and pharmacological point of view. The suitability of the autosolex for use in hospital radiopharmacy was shown by using the 99mTcO4- to prepare various 99mTc-RP and using these 99mTc-RP for scintigraphic imaging in patients.

A compact solvent extraction based 99Mo/99mTc 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.

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.

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).

A computerized compact module for separation of (99m)Tc-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.

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.