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PEMBUATAN DAN ANALISIS FISIKO-KIMIA RADIOISOTOP SKANDIUM-47 (47Sc) DARI BAHAN SASARAN TITANIUM OKSIDA ALAM Duyeh Setiawan; Titin Sri Mulyati
Jurnal Sains dan Teknologi Nuklir Indonesia (Indonesian Journal of Nuclear Science and Technology) Vol 16, No 2 (2015): Agustus 2015
Publisher : BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (350.612 KB) | DOI: 10.17146/jstni.2015.16.2.2379

Abstract

Radioisotop skandium-47 (47Sc) memiliki waktu paruh 3,35 hari, pemancar energi beta, Eβmax 0,441 MeV (68 %) dan 0,601 MeV (32 %), serta pemancar energi gamma, Eγ 159 keV (68 %). Radioisotop 47Sc dihasilkan oleh iradiasi neutron cepat dari target titanium berdasarkan reaksi inti 47Ti (n, p) 47Sc. Metode pemisahan 47Sc menggunakan cara kromatografi kolom dengan matriks Dowex AG 50W-x4 dalam bentuk kation (H+), selanjutnya 47Sc dielusi dengan HCl 4 M. Radioisotop 47Sc digunakan dalam bidang kedokteran nuklir untuk radioterapi dengan metode pencitraan. Karakteristik fisiko-kimia suatu sediaan radioisotop mempunyai peranan penting dalam penyebaran dan penimbunan di dalam tubuh. Oleh karena itu, untuk menjamin keberhasilan penggunaan sediaan radioisotop 47Sc perlu dilakukan analisis fisiko-kimia yang meliputi kejernihan, pH, kemurnian radionuklida dan radiokimia serta stabilitasnya pada penyimpanan. Hasil penelitian menunjukkan bahwa radio-isotop 47Sc berupa larutan jernih dengan rumus kimia 47Sccl3, memiliki pH 2, konsentrasi radioaktivitas 1,086 ± 0,0314 mCi/mL, aktivitas jenis 2,60 mCi/mg Ti (End of Irradiation = EOI), kemurnian radionuklida lebih dari 98,5 %, kemurnian radiokimia 95,22 ± 0,83 % dan masih stabil selama 5 hari disimpan di temperatur kamar. Radioisotop 47Sc yang diperoleh memiliki karakteristik fisiko-kimia untuk digunakan dalam pengembangan radiofarmaka sebagai sediaan radioterapi. PREPARATION AND PHYSICO-CHEMICAL ANALYSIS OF RADIOISOTOPE SCANDIUM-47 (47Sc) FROM NATURAL TITANIUM OXIDE MATERIAL TARGET. Radioisotopes scandium-47 (47Sc) has a half-life of 3.35 day, the energy beta transmitter Eβmax of 0.441 MeV (68 %) and 0.601 MeV (32 %), as well as gamma energy transmitter, Eγ 159 keV (68 %). Radioisotope 47Sc is produced by fast neutron irradiation of the titanium targets based on nuclear reaction 47Ti (n,p) 47Sc. Separation methods of 47Sc was done using chromatography column with a matrix of Dowex AG 50W-x4 in a cation (H+) form, and 47Sc was eluted with 4 M HCl. Radioisotope 47Sc is used in nuclear medicine for radiotherapy with imaging methods. The physico-chemical characteristics of a radioisotope has an important role in the biodistribution and bioaccumulation in the body. Therefore, in order to assure the success of usage of radioisotope 47Sc, of physico-chemical characteristic is need to be analyzed which includes clarity of solution, pH, purity of radionuclide and radiochemical, stability in the storage. The results showed that the radioisotope 47Sc was a clear solution with a chemical formula of 47ScCl3, has pH of 2 with the concentration of radioactivity 1,086 ± 0,0314 mCi/mL, specific activity of 2.60 mCi/mg Ti (End of Irradiation = EOI), the radionuclide purity more than 98.50 %, radiochemical purity 95,22 ± 0,83 % and  stable after 5 days storage in room temperature. Radioisotope 47Sc that was produced has the ideal physico-chemical characteristics and can be used for the radiopharmaceutical development especially for radiotherapy.
SYNTHESIS OF BUTHYL BROMIDE LABELED 82Br FOR LEAKAGE DETECTION APPLICATION IN INDUSTRIAL PIPELINE SYSTEM Ade Suherman; Titin Sri Mulyati; Iswahyudi Iswahyudi; Badra Sanditya Rattyananda; Dessy Cartika; Winda Putri Silpia; Mentik Hulupi; Duyeh Setiawan; Muhamad Basit Febrian
Jurnal Sains dan Teknologi Nuklir Indonesia (Indonesian Journal of Nuclear Science and Technology) Vol 20, No 2 (2019): Agustus 2019
Publisher : BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (286.836 KB) | DOI: 10.17146/jstni.2019.20.2.5597

Abstract

The detection of a leakage in an installation or pipeline in industrial complex is difficult to be done because related to security, safety, and operation condition. With expanded radioisotope application as a tracer in industry, hence a leakage in a pipe can be detected easily and qiuickly without needed excavation or stop the production process. The selection of radioisotope labeled compound as radiotracer should be examined carefully to determine the appropriate and well mixed radiotracer with the material passing through the pipeline system. Radioisotope labeled compound butyl bromide-82 (C4H982Br) as a radiotracer can be synthesized by reacting K82Br with 1-butanol and sulphuric acid (H2SO4) as a catalyst. The experiment result shows that synthesized C4H982Br by composition of 15 mL K82Br solution (0.1 gr/mL KBr) and 10 mL 1-butanol gave the highest percentage of reactions amount 50,00% and 40,95%. Characterization by FTIR showed that the product has absorption band for C-Br at 514,99-738,74 cm-1. GCMS analysis showed the peak of C4H982Br together with other 7 peaks of impurities with 43.03% percentage of C4H982Br peak. In distribution coefficient determination of C4H982Br in the test solution from industry (ethylene dichloride), Kd value of 5,1350 was obtained and more than 98% C4H982Br distilled together with ethylene dichloride in 110°C distillation process whereas no radioactivity detected in distillation flask if K82Br was used. Based on these results, C4H982Br is suitable to be applied as radiotracer for leakage detection in pipeline system with organic compounds as passing liquid including ethylene dichloride.
Performance of 113Sn/113mIn Generator Prototype based on Zirconium Oxide for Radiotracer Applications in Industry Duyeh Setiawan; Ade Suherman; Titin Sri Mulyati
Jurnal Sains dan Teknologi Nuklir Indonesia (Indonesian Journal of Nuclear Science and Technology) Vol 22, No 1 (2021): February 2021
Publisher : BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/jstni.2021.22.1.6486

Abstract

This paper describes the results of the research of 113Sn/113m radioisotope generator at the Centre for Applied Nuclear Science and Technology - National Nuclear Energy Agency with the aim of making radioisotope generator for radiotracer applications in industry. This research discussed the need of short half-life radiotracer for several material phases such as oil, water, gas and solid. Based on desired physical properties of half life, radiation type and energy, 113mIn radioisotope was selected. Thus, the performance of 113Sn/113mIn generator prototype based on zirconium oxide was determined for this purpose. The final product specification in the form of 113mInCl3 is clear solution, pH 2 with obtained yield of 95 %, radionuclide purity of 95 % and radiochemical purity of 95 %.