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PEMUNGUTAN ULANG BAHAN SASARAN TL-203 MELALUI PEMBENTUKAN TL-203(OH)3 DAN 2TL-2O3 Soenarjo, Sunarhadijoso; Triyanto, Triyanto; Yuliana, Nina; Indriani, Dian
Jurnal Radioisotop dan Radiofarmaka Vol 4, No 1,2 (2001): JURNAL PRR 2001
Publisher : Jurnal Radioisotop dan Radiofarmaka

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PEMUNGUTAN ULANG BAHAN SASARAN 203Tl MELALUI PEMBENTUKAN 203Tl(OH)3 DAN 203Tl2O3. Bahan sasaran 203Tl pengkayaan Tlnggi dipergunakan sebagai target pada produksi radioisotop medis 201Tl melalui reaksi inTl 203Tl(p,3n) 20lPb ?> 201Tl. Karena 203Tl pengkayaan Tlnggi sangat mahal dan Tldak selalu mudah didapatkan, maka diperlukan teknologi pemungutan ulang 203Tl dari limbah proses radioisotop 201Tl. Teknik pemungutan ulang yang menghasilkan bentuk akhir 203Tl2O3 padat menjadi pilihan karena 203Tl2O3 mempunyai kestabilan yang baik dan merupakan bahan dasar penyiapan sasaran. Rendemen pemungutan ulang juga mudah ditentukan karena hasil akhir berupa zat padat yang mudah diTlmbang. Proses pemungutan ulang mencakup 4 tahapan penTlng yaitu pemisahan 203Tl sebagai 203Tl(I) dari matriks limbah, oksidasi 203Tl(I) menjadi 203Tl(III) menggunakan Br2 dalam CHCl3, pengendapan 203Tl(III) sebagai 203Tl(OH)3 dan pengubahan 203Tl(OH)3 menjadi 203Tl2O3. Rendemen pemungutan ulang ditentukan dengan membandingkan berat 203Tl2O3 yang diperoleh terhadap kandungan 203Tl dalam limbah yang ditentukan terlebih dahulu dengan spektrofotometri UV pada panjang gelombang 214 nm. Pada proses simulasi menggunakan Tl alam didapatkan rata-rata rendemen pemungutan ulang sebesar (80,65 ± 1,05) % berdasarkan pada total Tl yang digunakan dalam simulasi limbah proses 201Tl. Penerapan proses pemungutan ulang dari campuran limbah pra-iradiasi dan limbah pasca-iradiasi menghasilkan rendemen pemungutan ulang 203Tl sebesar 81,02 %. CHEMICAL RECOVERY OF 203Tl TARGET THROUGH PRECIPITATION OF 203Tl(OH)3 AND CONVERSION TO 203Tl2O3. High-enriched 203Tl is preferably used as target material for production of 201Tl medical radioisotope through nuclear reaction of 203Tl(p,3n) 20lPb ?> 201Tl. Because of the high cost and limited availability of high-enriched 203Tl, a chemical recovery, of 203Tl from 201Tl processing waste is necessary. Method of chemical recovery giving final product as 203Tl2O3 is chosen due to reasons that 203Tl2O3 is stable and suitable for target preparation and the efficiency of the recovery can be easily determined. The chemical recovery covers the separation of 203Tl as 203Tl(I) from the waste matrices, oxidation of 203Tl(I), precipitation of the oxidized 203Tl as its hydroxide and conversion of the hydroxide to 203Tl2O3. The yield of the recovery was determined by comparing the weight of the resulting 203Tl2O3 to the initial 203Tl content that was previously measured by means of UV-spectrophotometry at a wavelength of 214 nm.From simulation processes using natural thallium, instead of enriched 203Tl, an average-recovery yield was found to be (80.65 ± 1.05)% based on the initial thallium used. The recovery of 203Tl from a mixture of pre and post irradiation wastes of 201Tl has been carried out giving a recovery yield of 81.02 %.
STUDY ON SEPARATION OF 137CS FROM 235U FISSION PROCESS WASTE. UTILIZATION OF SILICA GEL-SUPPORTED FERROCYANIDE COMPLEX SALT FOR 137CS PICKING Soenarjo, Sunarhadijoso; Pujianto, Anung
Jurnal Teknologi Bahan Nuklir Vol 4, No 1 (2008): Januari 2008
Publisher : PTBN - BATAN

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ABSTRACT STUDY ON SEPARATION OF 137Cs FROM 235U FISSION PROCESS WASTE. UTILIZATION OF SILICA GEL-SUPPORTED FERROCYANIDE COMPLEX SALT FOR 137Cs PICKING. In connection with the potential domestic demand especially in the fields of industry and nuclear medicine, the separation of 137Cs from 235U fission process waste is to be of interest although its economic value could be a polemic. A preliminary study on the separation of 137Cs from the 235U fission process waste generated in the production of 99Mo in P.T. BATAN Teknologi, Serpong, was performed through experiments on 137Cs picking from sample solution of the radioactive fission waste (RFW). The presented study is aimed to gain experimental data supporting utilization of the matrix of silica gel-supported ferrocyanide complex salt for the separation of 137Cs from RFW. Subsequent step would be the recovery and purification of 137Cs as part of production technology of 137Cs. The RFW sample was batch-treated with the matrix of silica gel-supported ferrocyanide complex salt which was synthesized from silica gel, potassium hexacyanoferrate(II) and copper(II) chloride. The binding of radioisotopes in RFW on the matrix was observed by y-spectrometry of the RFW solution before and during the treatment. The results showed that approximately 85% of 137Cs could be picked from the RFW sample into the matrix. Less amount of 95Zr and 95Nb was bound into the matrix. 103Ru was slightly bound into the matrix whereas 141/144Ce and 129mTe were not. It was observed that by using 0.2 and 0.4 g of matrix for 10 ml of RFW, the amount of matrix influenced the binding quantity of 95Zr and 95Nb but not that of 137Cs. FREE TERMS: Separation of 137Cs, 235U fission process, Ferrocyanide complex salt, Radioactive fission waste (RFW), y-spectrometry ABSTRAK STUDI PEMISAHAN 137Cs DARI LIMBAH PROSES FISI 235U. PENGGUNAAN MATRIK SILIKA GEL-GARAM KOMPLEK FEROSIANIDA UNTUK PEMUNGUTAN 137Cs. Berkaitan dengan potensi kebutuhan domestik, terutama di bidang industri dan kedokteran nuklir, pemisahan 137Cs dari limbah proses fisi 235U menjadi hal yang menarik walaupun mungkin masih menjadi perdebatan apakah ekonomis atau tidak. Sebagai studi awal untuk pemisahan 137Cs dari limbah proses fisi 235U, telah dilakukan percobaan pemungutan 137Cs dari cuplikan limbah radioaktif proses fisi (radioactive fission waste, RFW) yang dihasilkan pada proses produksi 99Mo di P.T. BATAN Teknologi, Serpong. Percobaan ini bertujuan untuk mendapatkan data eksperiment yang mendukung penggunaan matrik silika gel-garam komplek ferosianida untuk memisahkan 137Cs dari RFW. Tahapan selanjutnya adalah pemungutan dan pemurnian 137Cs sebagai bagian dari teknologi produksi 137Cs. Cuplikan RFW diperlakukan dalam proses batch dengan matrik silika gel-garam komplek ferosianida yang dibuat dari silika gel, kalium heksasianoferat(II) dan tembaga(II) klorida. Pengikatan radioisotop dalam RFW pada matrik diamati dengan teknik spektrometri-? terhadap larutan RFW sebelum dan selama perlakuan dengan matrik. Hasil percobaan menunjukkan bahwa sekitar 85% 137Cs dapat terambil dari larutan RFW dan terikat pada matrik. Dalam kuantitas yang lebih rendah, radionuklida 95Zr dan 95Nb juga dapat terikat pada matrik. Radionuklida 103Ru terikat pada matrik dalam jumlah yang relatif kecil sedangkan radionuklida 141/144Ce dan 129mTe tidak terikat. Dengan menggunakan 0,2 dan 0,4 gram matrik untuk perlakuan terhadap 10 ml larutan RFW, dapat diamati adanya pengaruh jumlah matrik terhadap kuantisasi pengikatan 95Zr dan 95Nb tetapi tidak terhadap pengikatan 137Cs. KATA KUNCI: Pemisahan 137Cs, Proses fisi 235U, Garam komplek ferosianida, Limbah radioaktif proses fisi (RFW), Spektrometri-
SIMULATIONS ON NICKEL TARGET PREPARATION AND SEPARATION OF NI(II)- Soenarjo, Sunarhadijoso; Rahman, Wira; Sriyono, Sriyono; Triyanto, Triyanto
GANENDRA Majalah IPTEK Nuklir Volume 14 Nomor 1 Januari 2011
Publisher : PSTA BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (211.268 KB) | DOI: 10.17146/gnd.2011.14.1.26

Abstract

SIMULATIONS ON NICKEL TARGET PREPARATION AND SEPARATION.OF Ni(II)-Cu(II) MATRIX FORPRODUCTION OF RADIOISOTOPE64Cu64Ni (p,n) 64  and retained on the column while the nickel was kept in the form of Ni2+  2+  2+ and CuCl while the nickel was totally in the form of Ni2+  while the nickel was found as both Ni2+ and NiCl   while the nickel was mostly in the form of Ni2+. The retained CuCl was then changed back into Cu2+     Keywords 64 Cu, Anion exchange chromatography.: Nickel target preparation, Radioisotope Cu-64, Separation of Ni(II)-Cu(II) matrix, Nuclear reaction of 64Ni(p,n) cation form andeluted out the column by using HCl 0.05 M. The 42? 4 2?.The best condition of separation was in HCl 8 M in which the radioactive copper was mostly in the form of CuCl 42? 42? . In the condition ofHCl 9 M, the radioactive copper was mostly in the form of CuCl 42? cation. It was found that the electroplating result from the acidic solution was more satisfied than that from the basic solution. By conditioning the matrix solution at HCl 6 M, the radioactivecopper was found in the forms of Cucation and eluted off from the column. The retained radioactive copper was then eluted out the column in the condition of dilute HCl changingback the copper anion complex into Cu42? Cu. The nickel target preparation was performed by means of electroplating method using acidic solution of nickel chloride - boric acid mixture and basic solution of nickel sulphate ? nickel chloridemixture on a silver- surfaced-target holder. The simulated solution of Ni(II) ? Cu(II) matrix was considered as thesolution of post-proton-irradiated nickel target containing both irradiated nickel and radioactive copper, but in thepresented work the proton irradiation of nickel target was omitted, while the radioactive copper was originallyobtained from neutron irradiation of CuO target. The separation of radioactive copper from the nickel target matrixwas based on anion exchange column chromatography in which the radiocopper was conditioned to form anioncomplex CuClg-spectrometric analysis showed a single strong peak at 511 keVwhich is in accord to g-annihilation peak coming from positron decay of Cu-64, and a very weak peak at 1346 keVwhich is in accord to g-ray of Cu-64.. The simulations on Nickel target preparation and separation of Ni(II)- Cu(II) matrix has been carried out as a preliminary study for production of medical radioisotope Cu-64 based onnuclear reaction of
RADIONUCLIDIC SEPARATION OF RADIOACTIVE INDIUM FOR MEDICAL AND BIOLOGICAL RESEARCH APPLICATIONS FROM TARGET MATRIX BASED ON NUCLEAR REACTION OF NATCD (N,γ) 115CD 􀃎 115MIN Soenarjo, Sunarhadijoso; Wisnukaton, Kadarisman; ., Sriyono; ., Abidin; ., Herlina
Jurnal Ilmiah Aplikasi Isotop dan Radiasi Vol 5, No 2 (2009): Desember 2009
Publisher : BATAN

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Abstract

Radioisotope 115mIn has been considered to be a very potential radioisotope formedical purposes and biological researches. Its physical properties are comparable to those of the radioisotope 99mTc. Although 115mIn is very potential for application innuclear medicine and biological researches, it is not widely explored for domestic use due to domestic limitations on its production technology. Accordingly, the objective of the present works is to master a production processing technology of 115mIn for medical and biological research applications. As the daughter of 115Cd, 115mIn is produced by neutron activation on cadmium target followed by separation in a radioisotope generator based on nuclear reaction of 114Cd (n,?) 115Cd ? 115mIn. In thisstudy, natural CdO was used as a target while the irradiation was carried out in the G.A. Siwabessy reactor. The separation of radioisotope 115mIn from the irradiated target was carried out by means of solvent extraction and anion exchange columnchromatography. In terms of solvent extraction, the post-irradiated target solution was extracted using two extractants namely 8-hydroxy-quinoline in chloroform and 2-ethylhexyl-phosphate in toluene. The resulting radioindium(III)-organo-complexwas then stripped from the organic phase to release the radioisotope 115mIn. Meanwhile in anion exchange column chromatography, the cadmium fraction in the post-irradiated target solution was conditioned to form anion complex, CdI42-, which was then bound on AG 1X8 (Cl¯, 100 ? 200 mesh) resin column. The formed 115mIn, the daughter of 115Cd, in the form of 115mIn3+ was then eluted from the column using 0.05 M HCl. It was found that the radioactive indium obtained from the solventextraction using 8-hydroxyquinoline in chloroform was chemically contaminated by the extractant, while that obtained from the solvent extraction using 2-ethylhexylphosphate in toluene was significantly contaminated by 115Cd. The anion exchangecolumn chromatography was found to be the best method for separation of 115mIn from post-irradiated target solution because this method produced pure 115mIn. This was indicated by the resulting 115mIn fraction that gave a mono-energetic ?-rayspectrum peaking at 336 keV and a half-life of 4.486 hours which were related to 115mIn. The quantitative aspect which was regarded as a radioactivity of the produced 115Cd was found to give a fluctuated result. This result was suspected to be inflicted by irradiation parameters such as inaccuracy in irradiation time, thechanges of reactor power and neutron flux as well as inter-irradiation-position load, which varied from one irradiation to another irradiation.
PERBANDINGAN Cr-51 (III) DAN Cr-S1 (VI) ANORGANIK PADA HASIL IRADIASI Cr(CO)s DAN Cr(C5H7O2)3 DENGAN NEUTRON TERMAL Soenarjo, Sunarhadijoso; Adam, Said; Idris, Iswandi
Jurnal Kimia Terapan Indonesia Vol 2, No 1-2 (1992)
Publisher : Research Center for Chemistry - LIPI

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Abstract

Thermal neutron irradiation on Cr(CO)6 and Cr(C5H7O2)3 target compounds had been conducted with an average neutron flux of about 2.8 x 10(12) n.cm(-2).der(1). Both post-irradiated target compounds gave inorganic radiochromium in oxidation states of +3 and +6 which were separated by solvent extraction method: For radioactivity measurement, the trivalent species was separased from the hexavalent by hydroxide precipitation using K2CrO4 and Cr(NO)3. 9H2O carriers. The inorganic chromium content was chemically determined by spectrophotometric method without adding arry carriers. The activity of the trivalent inorganic chromium produced from Cr(CO)6 irradiation was higher than that of the hexavalent ones, but in the case of Cr(C5H7O2)3 irradiation; the activity of the hexavalent species was higher. In both cases, the specific activity of the trivalent species was higher than that of the hexavalent species. The specific activity of total inorganic chromium obtained from the irradiation of Cr(CO)6 was higher than that of Cr(C5H7O2)3.
RADIONUCLIDIC IMPURITIES IN PERTECHNETATE SOLUTION ELUTED FROM 99mTc-CHROMATOGRAPHIC GENERATOR LOADED WITH 99Mo-FISSION PRODUCT Soenarjo, Sunarhadijoso; Hardi Gunawan, Adang
Jurnal Kimia Terapan Indonesia Vol 7, No 1-2 (1997)
Publisher : Research Center for Chemistry - LIPI

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Abstract

Medical radioisotope of 99mTc was firstly produced in Indonesia through the nuclear reaction of 98Mo (n;y) 99Mo -> 99mTc. The separation of the resulting 99mTc from the postirradiated natural MoO3 was carried out by solvent extraction using methyl ethyl ketone. Instead of this method, Radioisotope Production Center (RPC) BATAN has routinely produced 99mTc-chromatographic generator loaded with 235U-flSsion-produced 99Mo to provide 99mTc. By such generator, 99mTc can be easily and repeatedly liberated in the form of pertechnetate by vacuum elution using saline solution. Some fission- produced radionuclides, however, potentially contaminate the pertechnetate fraction. Gamma spectrometric determination was carried out to evaluate the level of radionuclidic impurities contaminating 99mTc-pertechnetate solution eluted from 99mTc- chromatographic generator produced over a 10-month period in 1993 - 1994. Radioactivity yield of the resulting 99mTc was independent to the origin of the loaded 99Mo. The 99mTc-pertechnetatate fractions were frequently contaminated with 99Mo, 131I and 103Ru, but the contamination did not exceed maximum permissible level. The fluctuation of contamination level may be influenced by irradiation parameters and separation techniques applied to the production of the loaded 99Mo.
Studi Pengikatan Cadmium (II) pada Resin Anorganik Titanium Oksida dan Zirkonium Oksida Soenarjo, Sunarhadijoso; Martati, Titiek; Winanti, Murti; Wirasti, Atmi
JURNAL ILMU KEFARMASIAN INDONESIA Vol 3 No 1 (2005): JIFI
Publisher : Fakultas Farmasi Universitas Indonesia

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Immobilization of Cd-(II) in production technology of indium medical radioisotopes (lll In or 115min) by means of 112Cd (p, 2n) 11 In or 114Cd (n,y) 115Cd 115mIn nuclear reaction is an important step regarding to the quality of the produced radioactive indium and efficiency of the utilization of the enriched targets (112Cd or 114Cd). The phenomena and capacity of Cd-(II) immobilization were studied using two kinds of inorganic oxide resins, i.e. titanium oxide and zirconium oxide. The aim of this study is gaining experimental data supporting the utilization of the resins in the production of 111In and 115min. The immobilization of Cd-(II) was proceeded by treating standard solutions of Cd-(II) with the resins followed by measurement of Cd-(II) content of the treated solution before and after treatment. The separation of the treated Cd-(II) solution from the resin was performed by means of centrifugation, while the measurement of Cd-(II) was performed by means of Ultra Violet spectrophotometry. The ready-used titanium oxide from Merck did not bind Cd-(II) even after being treated with several kinds of activating media. Zirconium oxide synthesized from the reaction of ZrOCI,. 8H,O with a basic solution of NaOH was in the form of the hydrate compound formulated as ZrO. nH,O with the value of n=(1.7034 +0.0186) and showed a capability to bind Cd-(II) higher than that of zirconium oxide synthesized with a basic solution of NH,OH or of the ready-used ZrO2. In general the Cd-(II)-binding capacity of Zro, tends to decrease with increasing of the amount of Zro.
Kapasitas Pengikatan Iodida dan Iod pada Karbon Aktif Konvensional dan Terbrominasi Soenarjo, Sunarhadijoso; Tamat, Swasono R.; Saputra, Ade
JURNAL ILMU KEFARMASIAN INDONESIA Vol 3 No 2 (2005): JIFI
Publisher : Fakultas Farmasi Universitas Indonesia

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Related to the activities in a radioisotope processing laboratory at CDRR, BATAN, active carbon has a great potency and a role as adsorber for radioisotope immobilizing agent in process and installation systems. KAKEN Corporation, a CDRR's collaboration partner in Japan, has introduced a new type of active carbon namely KAKEN Brominated Active Carbon (KBAC) which is claimed to have an anionic exchange character. The presented experiment was thus performed to determine the iodide and iodine binding capacities of the KBAC resin as well as comparing to those of the conventional active carbon, which was also provided by KAKEN (KCAC). The experiment had been done using solutions of natural iodide and iodide - iodine mixtures. The binding capacities were determined by iodometry and iodatometry. In general the results obtained showed that the iodide-binding capacity of KBAC was higher than that of KCAC, but that the iodine-binding capacity of KCAC was higher. The KBAC showed that the binding capacity of iodide from iodide solutions was smaller than that from iodide-iodine samples, while the KCAC showed the reverse.
KARAKTERISTIK PEMISAHAN RADIOLUTESIUM-177/177MLU DAN RADIOITERBIUM-169/175YB PADA KOLOM RESIN LN-EICHROM Widyaningrum, Triani; Triyanto, Mr; Sarmini, Endang; Sholikhah, Umi Nur; Soenarjo, Sunarhadijoso
Jurnal Sains dan Teknologi Nuklir Indonesia Vol 16, No 1 (2015): Februari 2015
Publisher : BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (472.61 KB) | DOI: 10.17146/jstni.2015.16.1.2353

Abstract

ABSTRAK KARAKTERISTIK PEMISAHAN RADIOLUTESIUM- 177/177mLu DAN RADIOITERBIUM-169/175Yb PADA KOLOM RESIN LN-EICHROM. Radiolutesium-177Lu keradioaktifan jenis tinggi merupakan salah satu radiolantanida yang banyak digunakan untuk menangani berbagai kasus kanker, namun di Indonesia penggunaan radiofarmaka bertanda 177Lu belum dapat dijanjikan karena teknik produksi radioisotop primernya belum dikuasai. Prospek produksi 177Lu melalui reaksi inti 176Yb (n,g) 177Yb* à 177Lu* + ?? dipelajari melalui metode pemisahan matrik 177/177mLu-169/175Yb/176Yb dalam sistem kromatografi kolom resin LN-Eichrom. Profil fraksinasi dan karakteristik pemisahan dipelajari dengan pemeriksaan keradioaktifan dan analisis spektro-metri-g terhadap hasil elusi larutan sasaran pasca iradiasi. Bahan sasaran digunakan 176Yb2O3 alam dan 176Lu2O3 diperkaya. Hasil penelitian menunjukkan bahwa radiolutesium-177/177mLu dapat dipisahkan dari matrik radioiterbium-169/175Yb/natYb melalui sistem kromatografi kolom dengan fase diam resin LN-Eichrom dan fase gerak larutan HNO3, dengan konsentrasi antara 1,5 ? 4 M untuk mendapatkan pemisahan yang efektif, selektif dan kuantitatif. Reaksi inti 176Yb(n,g) 177Yb* à 177Lu + ?? merupakan model reaksi inti yang perlu dipertimbangkan walau-pun harus melibatkan tahapan pemisahan produk 177Lu dari matrik sasaran Yb pasca iradiasi. Prosedur pemisahan yang dilakukan masih perlu diperbaiki melalui pemilihan jenis dan konsentrasi fase gerak pengelusi yang lebih tepat. ABSTRACT SEPARATION CHARACTERISTIC OF RADIOLUTETIUM-177/177mLu AND RADIOY-TTERBIUM-169/175Yb ON LN-EICHROM RESIN COLUMN. High specific activity radiolutetium-177Lu is one of radiolanthanides that is widely used to handle variety of cancer cases, but in Indonesia the use of 177Lu-labeled-radiopharmaceutical can not be promised yet as the primary radioisotope production techniques have not been mastered.  The  prospect of 177Lu production based on the nuclear reaction of 176Yb (n,g) 177Yb * ® 177Lu * + ?? in the BATAN?s G.A. Siwabessy reactor was learned through the separation characteristics of 177/177mLu-169/175Yb /176Yb process matrices in the LN-Eichrom resin column chromatography. The separation and fractionation profiles were characterized by radioactivity measurement as well as g-spectrometric analysis of the eluting post-irradiated target solution. The target materials used were natural 176Yb2O3 and enriched 176Lu2O3. The results showed that radiolutetium-177/177mLu can be separated from the radioiterbium-169/175Yb/natYb matrix by column chromatography system with a stationary phase of LN-Eichrom resin using HNO3 solution as the mobile phase, but the concentration of HNO3 used is a critical variable, between 1.5 - 4 M, to obtain an effective separation selectively and quantitatively. The nuclear reaction of 176Yb (n,g) 177Yb* ® 177Lu + ?? using natural Yb2O3 is considered to be better, although it must involve 177Lu product separation stage from the post-irradiated natural Yb target matrix. The presented separation procedure still needs to be improved through the selection of the type and the concentration of the mobile phase used to gain more appropriate elution solvent.
RSG-GAS BASED RADIOISOTOPES AND SHARING PROGRAM FOR REGIONAL BACK UP SUPPLY Soenarjo, Sunarhadijoso; Tamat, Swasono Rahardjo; Suparman, Ibon; Purwadi, Bambang
Jurnal Radioisotop dan Radiofarmaka Vol 6, No 2 (2003): Jurnal PRR 2003
Publisher : Jurnal Radioisotop dan Radiofarmaka

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ABSTRAK.RADIOISOTOP  BERBASIS  RSG-GAS  DAN PROGRAM KERJASAMA DUKUNGAN PASOKAN  REGIONALSebagai lembaga yang memiliki fasilitas reaktor untuk produksi radioisotop, BATAN perlu meningkatkan efektifItas biaya operasi reaktor melalui pencapaian pendayagunaan secara simultan keseluruhan fasilitas iradiasi yang tersedia dengan didukung pemanfaatan maksimal produk radioisotop yang dihasilkan. Di sisi lain, kebutuhan radioisotop di lingkungan domestik masih jauh di bawah kapabilitas maksimal produksi, tetapi ada kalanya pemenuhan kebutuhan radioisotop harus ditunda karena penyesuaian dengan jadwal operasi reaktor. Kondisi seperti ini teIjadi juga pada kebanyakan negara anggotaRCA (Bantuan KeIjasama Regional) - IAEA. Karena itu suatu program keIjasama dukungan pasokansediaan radioisotop secara regional merupakan pemikiran positif untuk pencapaian efektifItas biaya operasireaktor serta kesinambungan layanan dan pemanfaatan produk radioisotop yang  dihasilkan. Berdasarkan data hasil kegiatan produksi radioisotop domestik selama ini, maka radioisotop 311,9~0, J53Sm,12l dan 32pABSTRACT.RSG-GAS BASED RADIOISOTOPES AND SHARING PROGRAM FOR REGIONALBACK UP SUPPLY.As the owner of  the reactors used for radioisotope production, BATAN needs toincrease the effectiveness of the reactor operation cost that can be achieved by simultaneously exploiting allthe existing irradiation facility, supported by full utilization of the radioisotopes produced. On the other hand,the domestic demand of radioisotopes is much lower than the production capability but sometimes the requestis compulsory to be suspended due to reactor operation schedule. As this condition is mostly similar to that ofseveral countries of RCA Member States, a sharing program for regional back up supply seems to be apositive thought to support expectation on the effectiveness of reactor operation cost and the continuity ofradioisotope product services as well as the utilization of radioisotopes produced. Based on radioactivityachieved in each production batch at the present, 131r, 9~0, J53Sm,12srand 32pradioisotopes may be offeredfor back up supply program. Due to consideration on conformity of user demands with reactor operation andradiochemical processing costs, the concept of back up supply program should performed first by means offull utilization of the available products and not by increasing reactor operation frequency. An informationand communication network system, therefore, is absolutely needed to support infonnation exchange between the radioisotope producer, members of back up supply program and radioisotope customers.Key words: Radioisotopes production and supply, Reactor irradiation facility.