A solvent extraction-based procedure for removal of 46Sc impurity from reactor produced [45Ca]CaCl2 for its potential use in bone pain palliation.

Calcium-45 [T½ = 163 d, Eß (max) = 0.3 MeV] is a pure ß- emitting radioisotope which can be envisaged for potential use in palliative care of pain due to skeletal metastases of primary cancer. During production of 45Ca in nuclear reactor via 44Ca (n,γ) 45Ca route, 46Sc is co-produced as a radionuclidic impurity. In this study, we have optimized a single-step solvent extraction procedure for complete removal of 46Sc impurity from [45Ca]CaCl2. The purified radiotracer was administered intravenously in normal Wistar rats and preferential bone uptake could be demonstrated by ex vivo biodistribution studies.

A simple and robust method for radiochemical separation of no-carrier-added 64Cu produced in a research reactor for radiopharmaceutical preparation.

Copper-64 is an excellent theranostic radiometal that is gaining renewed attention of the clinical community in the recent times. In order to meet the increasing demand of this radiometal, we have demonstrated the viability of its production via 64Zn (n,p) 64Cu reaction in a nuclear reactor. A semi-automated radiochemical separation module based on selective extraction of 64Cu as dithizonate complex was developed. The maximum available activity at the end of irradiation was ~ 700 MBq. The overall yield of 64Cu after the separation process was >85% and it could be obtained with ~12 GBq/µg specific activity, >99.9% radionuclidic purity and >98% radiochemical purity. The separated 64Cu could be utilized for preparation of a wide variety of radiopharmaceuticals.

An electrochemical approach for removal of radionuclidic contaminants of Eu from 153Sm for effective use in metastatic bone pain palliation.

INTRODUCTION: Thermal neutron activation of 152Sm [152Sm(n,γ)153Sm] using natural or isotopically enriched (by 152Sm) samarium target is the established route for production of 153Sm used for preparation of 153Sm-EDTMP for pain palliation in cancer patients with disseminated bone metastases. However, some long-lived radionuclidic contaminants of Eu, such as, 154Eu (t½=8.6y) are also produced during the target activation process. This leads to detectable amount of Eu radionuclidic contaminants in patients' skeleton even years after administration with therapeutic doses of 153Sm-EDTMP. Further, the presence of such contaminants in 153Sm raises concerns related to radioactive waste management. The aim of the present study was to develop and demonstrate a viable method for large-scale purification of 153Sm from radionuclidic contaminants of Eu. METHODS: A radiochemical separation procedure adopting electroamalgamation approach has been critically evaluated. The influence of different experimental parameters for the quantitative removal radionuclidic contaminants of Eu from 153Sm was investigated and optimized. The effectiveness of the method was demonstrated by purification of ~37 GBq of 153Sm in several batches. As a proof of concept, 153Sm-EDTMP was administered in normal Wistar rats and ex vivo γ-spectrometry of bone samples were carried out. RESULTS: After carrying out the electrolysis under the optimized conditions, the radionuclidic contaminants of Eu could not be detected in purified 153Sm solution by γ-spectrometry. The overall yield of 153Sm obtained after the purification process was >85%. The reliability of this approach was amply demonstrated in several batches, wherein the performance remained consistent. Ex vivo γ-spectrometry of bone samples of Wistar rats administered with 153Sm-EDTMP (prepared using electrochemically purified 153Sm) did not show photo peaks corresponding to radionuclidic contaminants of Eu. CONCLUSIONS: A viable electrochemical strategy for the large-scale purification of 153Sm from radionuclidic contaminants of Eu has been successfully developed and demonstrated.

Initial Clinical Experience with 68Ga-DOTA-NOC Prepared Using 68Ga from Nanoceria-polyacrylonitrile Composite Sorbent-based 68Ge/68Ga Generator and Freeze-dried DOTA-NOC Kits.

Somatostatin receptor positron emission tomography-computed tomography (PET/CT) with 68Ga-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA) peptides have become an indispensable part of disease assessment in patients with neuroendocrine tumors and forms the basis of personalized therapy with peptide receptor-based radionuclide therapy. With growing utilization of PET/CT in developing countries, availability of the indigenous GMP-certified 68Ge/68Ga generators is expected to further promote cost-effective molecular imaging service to the cancer patients. We present our initial clinical experience in 32 patients injected with 68Ga-DOTA-NOC prepared using 68Ga eluted from Bhabha Atomic Research Centre nanoceria-polyacrylonitrile sorbent-based 68Ge/68Ga generator and freeze-dried DOTA-NOC cold kits.

Mechanochemical synthesis of mesoporous tin oxide: a new generation nanosorbent for (68)Ge/(68)Ga generator technology.

The present article reports the synthesis and characterization of mesoporous tin oxide (MTO) nanoparticles by a solid-state mechanochemical route. The synthesized material was used as an advanced sorbent material for (68)Ge/(68)Ga radionuclide generator technology. Gallium-68 (t½ = 68 min) obtained from the (68)Ge/(68)Ga generator is an important diagnostic radioisotope which holds tremendous potential in the non-invasive monitoring of various diseases, including cancer, using positron emission tomography (PET). The crystallite size of the MTO nanoparticles was in the range of 6-12 nm with a large surface area of 265 ± 16 m(2) g(-1), while the mean pore radius was found to be 2.1 ± 0.6 nm. Determination of the zeta-potential of the MTO nanoparticles dispersed in solutions at different pH values aided in understanding the sorption and separation mechanisms, which were based on the surface charge developed on the nanosorbent. The sorption capacity observed under column-flow conditions was 85 ± 5 mg Ge per g of nanosorbent. A clinical-scale (68)Ge/(68)Ga generator (740 MBq) was developed using this nanosorbent. Gallium-68 could be regularly eluted from this generator over a prolonged period of 1 year with >70% elution yield and met all the requirements for clinical use. The suitability of (68)Ga obtained from it was evaluated in preclinical settings by the preparation of a (68)Ga-labeled peptide containing the arginine-glycine-aspartic acid (RGD) motif. To the best of our knowledge, this is the first report on the synthesis of MTO nanoparticles by a mechanochemical route which could be effectively utilized for the routine preparation of clinical-scale (68)Ge/(68)Ga generators. The promising results obtained in this study would facilitate greater implementation of mechanochemistry for the synthesis of nanosorbents for radionuclide generator technology since this method is simple, economical and convenient.

Synthesis and evaluation of a (68)Ga labeled folic acid derivative for targeting folate receptors.

Present work evaluates the potential of a newly synthesized (68)Ga-NOTA-folic acid conjugate for PET imaging of tumors over-expressing folate receptors (FRs). NOTA-folic acid conjugate was synthesized and characterized. It was radiolabeled with (68)Ga in ≥ 95% radiolabeling yields. In vitro cell binding studies showed a maximum cell uptake of 1.7±0.4% per million KB cells which was completely blocked on addition of cold folic acid showing specificity towards the FRs. However, further studies in tumor xenografts are warranted in order to assess the potential of (68)Ga-folic acid complex for imaging tumors over-expressing FRs.

Detailed evaluation of different (68)Ge/(68)Ga generators: an attempt toward achieving efficient (68)Ga radiopharmacy.

The present study is aimed at carrying out a comparative performance evaluation of different types of (68)Ge/(68)Ga generators to identify the best choice for use in (68)Ga-radiopharmacy. Over the 1 year period of evaluation, the elution yields from the CeO2-based and SiO2-based (68)Ge/(68) Ga generators remained almost consistent, in contrast to the sharp decrease observed in the elution yields from TiO2 and SnO2-based generators. The level of (68)Ge impurity in (68)Ga eluates from the CeO2 and SiO2-based (68)Ge/(68)Ga generator was always <10(-3)%, while this level increased from 10(-3)% to 10(-1)% in case of TiO2 and SnO2-based generators. The level of chemical impurities in (68)Ga eluates from CeO2 and SiO2-based (68)Ge/(68)Ga generators was negligibly low (<0.1 ppm) in contrast to the significantly higher level (1-20 ppm) of such impurities in eluates from other two generators. As demonstrated by radiolabeling studies carried out using DOTA-coupled dimeric cyclic RGD peptide derivative (DOTA-RGD2), CeO2-PAN and SiO2-based generators are directly amenable for radiopharmaceutical preparation, whereas the other generators can be only used after post-elution purification of (68)Ga eluates. Clinically relevant dose of (68)Ga-DOTA-RGD2 was prepared in a hospital radiopharmacy for non-invasive visualization of tumors in breast cancer patients using positron emission tomography.

Palliative care of bone pain due to skeletal metastases: Exploring newer avenues using neutron activated (45)Ca.

INTRODUCTION: With an objective to develop a cost-effective radiochemical formulation for palliation of pain due to skeletal metastases, we have demonstrated a viable method for large-scale production of (45)Ca (t½=163 days, Eßmax=0.3MeV) using moderate flux research reactor, its purification from radionuclidic impurities adopting electrochemical approach and preclinical evaluation of (45)CaCl2. METHODS: Irradiation parameters were optimized by theoretical calculations for production of (45)Ca with highest possible specific activity along with minimum radionuclidic impurity burden. Based on this, the radioisotope was produced in reactor by irradiation of isotopically enriched (98% in (44)Ca) CaO target at a thermal neutron flux of ~1 × 10(14) n.cm(-2).s(-1) for 4 months. Scandium-46 impurity co-produced along with (45)Ca was efficiently removed adopting an electrochemical separation approach. The bone specificity of (45)CaCl2 was established by in vitro studies involving its uptake in hydroxyapatite (HA) particles and also evaluating its biodistribution pattern over a period of 2 weeks after in vivo administration in Wistar rats. RESULTS: Thermal neutron irradiation of 100mg of enriched (98% in (44)Ca) CaO target followed by radiochemical processing and electrochemical purification procedure yielded ~37 GBq of (45)Ca with a specific activity of ~370 MBq/mg and radionuclidic purity>99.99%. The reliability and reproducibility of this approach were amply demonstrated by process demonstration in several batches. In vitro studies indicated significant uptake of (45)CaCl2 (up to 65%) in HA particles. In vivo biodistribution studies in Wistar rats showed specific skeletal accumulation (40-46%ID) with good retention over a period of 2 weeks. CONCLUSIONS: To the best of our knowledge, this is the first study on utilization of (45)CaCl2 in the context of nuclear medicine. The results obtained in this study hold promise and warrant further investigations for future translation of (45)CaCl2 to the clinics, thereby potentially enabling a cost-effective approach for metastatic bone pain palliation especially in developing countries.

Usefulness of nano-zirconia for purification and concentration of ¹²5I solution for medical applications.

This paper described the utility of tetragonal nano-zirconia (t-ZrO2) to remove trace levels of ¹³4Cs and ¹³7Cs contaminants from ¹²5I solution obtained from neutron irradiation of natural Xe target. A careful scrutiny of the adsorption parameters of t-ZrO2 was considered worthwhile investigating to arrive at the optimum conditions to perform the purification as well as concentration of ¹²5I solution. The procedure proposed herein provides ¹²5I of acceptable purity and radioactive concentration for clinical application. The compatibility of the separated product in the preparation of ¹²5I brachytherapy sources was found to be satisfactory.

Long-term evaluation of 'BARC 68Ge/68Ga generator' based on the nanoceria-polyacrylonitrile composite sorbent.

This article describes the long-term evaluation of a nanoceria-polyacrylonitrile (CeO2-PAN) composite sorbent-based (68)Ge/(68)Ga generator reported. This generator used the new CeO2-PAN composite sorbent for preparation of the (68)Ge/(68)Ga generator. Since this sorbent has not been previously evaluated, a thorough long-term evaluation of the performance of the generator is necessary to ensure its applicability for clinical practice. The performance of the generator was evaluated in terms of (68)Ga yield, (68)Ge breakthrough, radioactive concentration of the (68)Ga solution, and suitability of the (68)Ga for the preparation of (68)Ga-labeled tracers. The (68)Ge/(68)Ga generator was able to provide a (68)Ga activity with consistent yields (>70%) and having acceptable radionuclidic (<10(-4)% of (68)Ge breakthrough), radiochemical, and chemical purities for an extended period of time. The eluted (68)GaCl3 is useful for the majority of the (68)Ga complexation chemistry.

Preparation of clinical-scale 99Mo/99mTc column generator using neutron activated low specific activity 99Mo and nanocrystalline γ-Al2O3 as column matrix.

INTRODUCTION: Preparation of clinical-scale (99)Mo/(99m)Tc generator using (n,γ) activated low specific activity (99)Mo and nanocrystalline γ-Al(2)O(3) as a high capacity sorbent matrix is attempted. METHODS: Nanocrystalline γ-Al(2)O(3) was synthesized by 'solid state mechanochemical' reaction of aluminum nitrate with ammonium bicarbonate. Experimental parameters were optimized to effectively separate (99m)Tc from (99)Mo using this sorbent as the column matrix. The performance features of a 13 GBq (350 mCi) (99)Mo/(99m)Tc generator using this sorbent and (99)Mo produced by (n,γ) route having specific activity 12.9-18.5 GBq/g were evaluated for 10 days. RESULTS: The sorbent possessed the requisite selectivity for (99)Mo and demonstrated a maximum sorption capacity of 200 ± 5mg Mo/g, which is ~10 times higher than that of ordinary acidic alumina. The overall yield of (99m)Tc was >80%, with radionuclidic purity >99.99% and radiochemical purity >99%. The yield of (99m)Tc varied from 7.8 to 2.1 GBq in the eluate for the six days of operation of the generator. The radioactive concentration of (99m)Tc eluted was adequate for the formulation of radiopharmaceuticals. The performance of the generator remained consistent over an extended period of 10 days. The eluted (99m)Tc was suitable for the formulation of (99m)Tc-DMSA and (99m)Tc-EC resulting in high radiolabeling yields (>98%). CONCLUSION: The effectiveness of γ-Al(2)O(3) as a new generation sorbent in the development of clinically useful (99)Mo/(99m)Tc generator using low specific activity (99)Mo and yielding (99m)Tc with adequate radioactive concentration and high purity suitable for formulation of radiopharmaceuticals is demonstrated.

Ammonium molybdophosphate impregnated alumina microspheres as a new generation sorbent for chromatographic 137Cs/(137m)Ba generator.

Barium-137m availed from a (137)Cs/(137m)Ba generator is an attractive industrial radiotracer for liquid flow rate measurement owing to its short half-life (T(1/2)=2.5 min) and emission of highly energetic gamma radiation (E(γ)=661.7 keV). Ammonium molybdophosphate (AMP) impregnated alumina microspheres (AMP-Al(2)O(3)), an engineered form of sorbent material was synthesized by Gel Entrapment Technique (GET). The utility of the material in the preparation of (137)Cs/(137m)Ba generator was evaluated. Various experimental parameters were optimized for efficient retention of (137)Cs and selective elution of (137m)Ba. A chromatographic 0.74 GBq (20 mCi) (137)Cs/(137m)Ba generator was developed using this material and its performance was evaluated for 6 months. Barium-137m could be eluted from the generator using 0.1 M NH(4)NO(3)+0.5 M HNO(3) solution with >80% yield and with acceptable radionuclidic purity suitable for industrial radiotracer investigations.

Development of a 99Mo/99mTc generator using alumina microspheres for industrial radiotracer applications.

A chromatographic (99)Mo/(99m)Tc generator for industrial applications has been developed using alumina microspheres synthesized through sol-gel process to obtain (99m)Tc in both aqueous and non-aqueous media. The sorbent was mesoporous, mechanically strong and possessed high surface area. (99m)Tc could be eluted from generator system using either acetone or 0.9% NaCl solution with appreciably high yields and high radiochemical as well as radionuclidic purity. The facile, versatile generator provides an efficient way to access (99m)Tc at industrial sites for radiotracer applications.

Exploitation of nano alumina for the chromatographic separation of clinical grade 188Re from 188W: a renaissance of the 188W/188Re generator technology.

The (188)W/(188)Re generator using an acidic alumina column for chromatographic separation of (188)Re has remained the most popular procedure world over. The capacity of bulk alumina for taking up tungstate ions is limited (∼50 mg W/g) necessitating the use of very high specific activity (188)W (185-370 GBq/g), which can be produced only in very few high flux reactors available in the world. In this context, the use of high-capacity sorbents would not only mitigate the requirement of high specific activity (188)W but also facilitate easy access to (188)Re. A solid state mechanochemical approach to synthesize nanocrystalline γ-Al(2)O(3) possessing very high W-sorption capacity (500 mg W/g) was developed. The structural and other investigations of the material were carried out using X-ray diffraction (XRD), transmission electron microscopy (TEM), Brunauer Emmett Teller (BET) surface area analysis, thermogravimetric-differential thermal analysis (TG-DTA), and dynamic light scattering (DLS) techniques. The synthesized material had an average crystallite size of ∼5 nm and surface area of 252 ± 10 m(2)/g. Sorption characteristics such as distribution ratios (K(d)), capacity, breakthrough profile, and elution behavior were investigated to ensure quantitative uptake of (188)W and selective elution of (188)Re. A 11.1 GBq (300 mCi) (188)W/(188)Re generator was developed using nanocrystalline γ-Al(2)O(3), and its performance was evaluated for a period of 6 months. The overall yield of (188)Re was >80%, with >99.999% radionuclidic purity and >99% radiochemical purity. The eluted (188)Re possessed appreciably high radioactive concentration and was compatible for the preparation of (188)Re labeled radiopharmaceuticals.

Development of a nano-zirconia based 68Ge/68Ga generator for biomedical applications.

INTRODUCTION: Most of the commercially available (68)Ge/(68)Ga generator systems are not optimally designed for direct applications in a clinical context. We have developed a nano-zirconia based (68)Ge/(68)Ga generator system for accessing (68)Ga amenable for the preparation of radiopharmaceuticals. METHODS: Nano-zirconia was synthesized by the in situ reaction of zirconyl chloride with ammonium hydroxide in alkaline medium. The physical characteristics of the material were studied by various analytical techniques. A 740 MBq (20 mCi) (68)Ge/(68)Ga generator was developed using this sorbent and its performance was evaluated for a period of 1 year. The suitability of (68)Ga for labeling biomolecules was ascertained by labeling DOTA-TATE with (68)Ga. RESULTS: The material synthesized was nanocrystalline with average particle size of ~7 nm, pore-size of ~4 Å and a high surface area of 340±10 m(2) g(-1). (68)Ga could be regularly eluted from this generator in 0.01N HCl medium with an overall radiochemical yield >80% and with high radionuclidic (<10(-5)% of (68)Ge impurity) and chemical purity (<0.1 ppm of Zr, Fe and Mn ions). The compatibility of the product for preparation of (68)Ga-labeled DOTA-TATE under the optimized reaction conditions was found to be satisfactory in terms of high labeling yields (>99%). The generator gave a consistent performance with respect to the elution yield and purity of (68)Ga over a period of 1 year. CONCLUSIONS: The feasibility of preparing an efficient (68)Ge/(68)Ga generator which can directly be used for biomedical applications has been demonstrated.

Nanoceria-PAN composite-based advanced sorbent material: a major step forward in the field of clinical-grade 68Ge/68Ga generator.

The (68)Ge/(68)Ga generator has high potential for clinical positron emission tomography (PET) imaging. However, because of the unavailability of a suitable sorbent material, the commercially available (68)Ge/(68)Ga generators are not directly adaptable for the preparation of (68)Ga-labeled radiopharmaceuticals. In view of this, a new nanoceria-polyacrylonitrile (PAN) composite sorbent has been synthesized by decomposition of a cerium oxalate precursor to cerium oxide and its subsequent incorporation in PAN matrix for the development of a clinical grade (68)Ge/(68)Ga generator. The X-ray diffraction (XRD) studies and BET nitrogen adsorption technique revealed that nanocrystalline ceria had an average particle size of approximately 10 nm, surface area of 72 +/- 3 m(2)/g and an average pore size of 3.8 +/- 0.1 A. Investigation of the distribution ratio (K(d)) values for the prepared sorbent in 0.01 N HCl medium revealed the suitability of the sorbent for the quantitative retention of (68)Ge and efficient elution of clinical grade (68)Ga. A 370 MBq (10 mCi) (68)Ge/(68)Ga chromatographic generator was developed using this sorbent. (68)Ga could be regularly eluted from this generator with >80% elution yield. The eluted (68)Ga possess high radionuclidic purity (<1 x 10(-5)% of (68)Ge impurity), chemical purity (<0.1 ppm of Ce, Fe and Mn ions) and was amenable for the preparation of (68)Ga-labeled radiopharmaceuticals. The generator gave a consistent performance with respect to the elution yield and purity of (68)Ga over an extended period of 7 months.

Fabrication of cesium-137 brachytherapy sources using vitrification technology.

137Cs source in solid matrix encapsulated in stainless-steel at MBq (mCi) levels are widely used as brachytherapy sources for the treatment of carcinoma of cervix uteri. This article describes the large-scale preparation of such sources. The process of fabrication includes vitrification of 137Cs-sodium borosilicate glass, its transformation into spheres of 5-6 mm diameter, casting of glass spheres into a cylinder of 1.5 mm (varphi) x 80 mm (l) in a platinum mould, cutting of the moulds into 5-mm-long pieces, silver coating on the sources, and finally, encapsulation in stainless steel capsules. Development of safety precautions used to trap 137Cs escaping during borosilicate glass preparation is also described. The leach rates of the radioactive sources prepared by the above technology were within permissible limits, and the sources could be used for encapsulation in stainless steel capsules and supplied for brachytherapy applications. This development was aimed at promoting the potential utility of 137Cs-brachytherapy sources in the country and reducing the user's reliance on imported sources. Since its development, more than 1000 such sources have been made by using 4.66 TBq(126 Ci) of 137Cs.

Recovery and purification of 4.66 TBq(126Ci) of 137Cs from a 20-year-old spent sealed source.

An attempt to recover and purify 4.66 TBq(126Ci) of (137)Cs as solution from a 20-year-old doubly encapsulated sealed source containing (137)CsCl was made successfully. The primary capsule was first cut open to retrieve the secondary capsule. The top end of the secondary SS capsule was then cut to render it as an open cylinder and the (137)CsCl pellet inside was dissolved in small aliquots of water. Each aliquot of the (137)CsCl solution was dispensed into glass vials, sealed, assayed and stored. Barium formed during the radioactive decay of (137)Cs was removed by carbonate precipitation. Chloride was removed by heating with HNO(3) and evaporation. About 91.8% of the total activity could be recovered and used subsequently for making brachytherapy sources.

Polymer embedded nanocrystalline titania sorbent for 99Mo-99mTc generator.

A new sorbent material, polymer embedded nano crystalline titania (Titanium Polymer-TiP) has been developed, from titanium (IV) chloride and isopropyl alcohol, for the adsorption of 99Mo, which is a precursor to 99mTc, a workhorse in radio-pharmaceuticals. The infrared absorption spectra of the TiP showed peaks corresponding to Ti-O groups. X-ray diffraction pattern of the adsorbent corresponded to rutile TiO2. The surface area of this polymer was 30 m2/g with an average pore size of 40 nm. The average crystallite size of TiO2, embedded in polymer, was found to be 5 nm. TEM micrograph of the adsorbent revealed the network of polymer with dispersed titania phase. Potential of this adsorbent for the preparation of 99Mo-99mTc generator has been explored. 99Mo could be adsorbed on to the adsorbent column containing TiP at pH 1 from which 99mTC could be eluted with normal (0.9%) saline solution with an elution yield of approximately 80%. The quality of the 99mTcO4 obtained was in accordance with the international specifications applicable for radiopharmaceutical use. A process demonstration run was carried out with 1.1 GBq (30 mCi) 99Mo activity level making use of the above adsorbent and consistent results were obtained over a period of one week, which is generally the shelf life of 99MO-99mTC generator.

Bioevaluation studies of 32P incorporated mould brachytherapy sources for potential application in treatment of superficial tumors.

AIM: To prepare 32P-based user-friendly mould brachytherpy sources for the treatment of superficial tumors. METHODS: 32P as orthophosphoric acid was adsorbed on 15-25 mm (diameter) circular sheets of cellulose-based adsorbent paper to prepare sources containing approximately 37-74 MBq of 32P per cm of strip. The sources were immobilized between plastic sheets of 40 microm thickness. Autoradiography studies were carried out to determine the uniformity of 32P deposition in the source. Dosimetric evaluation of the sources was also carried out. Bioevaluation studies were carried out in C57BL6 mice bearing melanoma using 37-74 MBq sources. RESULTS: Cellulose-based sources containing 37-74 MBq of 32P per cm2 could be prepared from which no radioactivity leakage could be detected in water or saline. Autoradiography studies revealed 32P to be uniformly distributed in these sources. Dosimetric evaluation showed that the contact dose imparted was 10 Gy/h, sufficient for treatment of superficial tumors. In mice bearing melanoma, complete tumor regression could be achieved with two applications of 37-74 MBq sources, at an interval of 3 days. Histopathological examination of the skin tissue from the treated area proved the absence of tumor as compared with the controls. CONCLUSION: Preparation of P sources of various shapes and sizes (based on the tumor size) having uniform 32P activity distribution could be achieved. Efficacy of these sources in treating melanoma tumors could be established in the animal model.