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Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Published by BRIN Publishing
ISSN : 08524777     EISSN : 25280473     DOI : https://doi.org/10.55981/urania
Core Subject :
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir is a scientific publication platform that presents experimental research and development studies on the nuclear fuel cycle, covering radiation physics and chemistry, reactor engineering, nuclear safety and security, nuclear forensics, and radioactive waste management
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Articles 183 Documents
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PEMBUATAN SENYAWA BERTANDA 131I-MIBG DENGAN PENINGKATAN DOSIS TERAPI Maskur Maskur; Chairuman; Yono Sugiharto; Khoirunnisa Fauziah Asyikin; Triningsih; Sulaiman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.1.6358

Abstract

Nowadays, the labeled compound of 131I-MIBG has been successfully produced in Indonesia. However, the concentration of radioactivity and dose of radioactivity per vial is still small when compared to commercial products from abroad available in the market. Currently, the dose of radioactivity achieved is 25 mCi/vial with a concentration of 5 mCi/mL. Problems arise if the radioactivity and dose of radioactivity per bottle are enlarged, where the product becomes unstable and breaks down because of the effects of radiolysis. The generally global commercial products contain a dose of radioactivity = 100 mCi/vial with a concentration of 10 mCi/mL. By this research efforts have been made to increase the concentration and dose of radioactivity per bottle. The method used was by increasing the amount of benzyl alcohol from 9 μL/mL to 10 μL/mL of 131I-MIBG solution. Other than that, the addition of phosphate buffer which was originally added before the purification process was switched to after the purification process. The result of labeling was 131I-MIBG 153 - 254 mCi with process yield between 33.60 and 51.94%. The radioactivity concentration was quite large, i.e., between 9.80 to 25.40 mCi/mL with radioactivity doses ranging from 100 to 222 mCi/vial. However, to adjust to market needs, the I-MIBG of Batan product was then packaged in 100 mCi/vial with a concentration of 5-13 mCi/mL.This workaround shows that the synthesis of labeled compound of 131I-MIBG for scale up therapeutic dose has been successfully carried out.
VALIDATION OF SRAC CODE SYSTEM FOR NEUTRONIC PARAMETERS CALCULATION OF THE PWR MOX/UO2 CORE BENCHMARK Wahid Luthfi; Surian Pinem
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.1.6238

Abstract

Determination of neutronic parameter value is an important part in determining reactor safety, so accurate calculation results can be obtained. This study is focused on the validation of SRAC code system in the calculation of neutronic parameters value of a PWR (Pressurized Water Reactor) reactor core. MOX/UO2 Core Benchmark was choose because it is used by several researchers as a reference core for code validation in the determination of neutronic parameters of a reactor core. The neutronic parameters calculated include critical boron concentration, delayed neutron fraction and Power Peaking Factor (PPF), and its distribution in axial and radial directions. When compared with reference data, the calculation results of the critical boron concentration value show that there is a difference of 22.5 ppm on SRAC code system. Meanwhile, differences in power per fuel element (assembly power error) value of power-weighted error (PWE) and error-weighted error (EWE) is 2.93% and 3.94%, respectively. Maximum difference between PPF value in axial direction with reference reaches a value of 4.57%. SRAC calculation results also show consistency with the calculation results of other program packages or code. Results of this study indicate that SRAC code system is still quite accurate for the calculation of neutronic parameters of PWR reactor core benchmark. Therefore, SRAC code system can be used to calculate neutronic parameters of PWR reactor core, especially when using MOX (mixed oxide) fuel.
PENENTUAN UMUR YELLOW CAKE SECARA RADIOKRONOMETRI Erlina Noerpitasari; Syamsul Fatimah; Iis Haryati; Yanlinastuti Yanlinastuti; Jan Setiawan; Boybul Boybul; Arif Nugroho
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.2.6057

Abstract

YELLOW CAKE AGE DETERMINATION BY RADIOCHRONOMETRY. Radiochronometric dating of yellow cake has been carried out. The aim of this study was to obtain data on the age of yellow cake for compiling a database for nuclear forensics library. The radiochronometer used was228Th/232Th. 232Th is thorium which naturally co-exists in yellow cake and will reach a decay equilibrium to 228Th. In the yellow cake production process, however, the equilibrium is disturbed so that the "time" will be zeroed again. The test samples were yellow cake from Cogema, from PTBGN mining in Kalan, and by-products from Petrokimia Gresik. The steps of processes include dissolving the sample, separating thorium from uranium by cation exchange chromatography using dowex 50W-X8 resin, electrodeposition of thorium and isotopes analysis using an alpha spectrometer. Age determination was performed by calculating the 228Th/232Th activity ratio and decay formula iteration. Based on the analysis of the 228Th and 232Th isotopes using an alpha spectrometer and the results of age calculations, the age of Cogema yellow cake was not able to be determined because it does not contain 232Th. The age of Petrokimia Gresik yellow cake is 9.90 years with a bias of 20.35 years and the age of PTBGN yellow cake is 12.85 years with a bias of more than 20.15 years when compared to the known estimated production time. The bias obtained is higher when compared to the results determined by previous researchers by the same method,i.e., a bias of less than 10 years.
PENGARUH ARAH PENGEROLAN DAN SHEAR CUTTING TERHADAP KARAKTER PADUAN AlMg2 SEBAGAI MATERIAL CLADDING BAHAN BAKAR REAKTOR RISET Sungkono Sungkono; Sri Ismarwanti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6090

Abstract

THE INFLUENCE OF ROLLING DIRECTION AND SHEAR CUTTING ON THE CHARACTERISTICS OF AlMg2 AS FUEL CLADDING MATERIAL FOR RESEARCH REACTOR FUEL. Gradual rolling process of AlMg2 alloy has been carried out to obtain a certain thickness with final mechanical behavior suitable for fuel cladding of research reactor. The post-irradiation testing of fuel element used mini specimens to minimize radioactive waste. The purpose of this study was to study the influence of rolling direction and shear cutting on AlMg2 alloy characteristics. The tests performed include tensile, metalographic and hardness tests after rolling. The results show that AlMg2 alloy rolled by same direction has lower tensile strength and hardness and is more ductile than that rolled by transverse direction. The structure of AlMg2 alloy after hot rolling consists of  (Al) phase with precipitates distributed along the elongated grains in the rolling direction. Shear cutting affects the width and hardness of the deformation area on the cut surface of the specimens.
MICROSTRUCTURAL EVOLUTION AND COARSENING KINETICS OF B2-(Fe,Ni)Al IN FERITIC STEEL OF Fe-14Ni-9Al-7,5Cr-1Mo AT HIGH TEMPERATURES. Eddy Agus Basuki; Syaiful Bachri; Djoko Hadi Prajitno
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6043

Abstract

MICROSTRUCTURAL EVOLUTION AND COARSENING KINETICS OF B2-(Fe,Ni)Al IN FERRITIC STEEL OF Fe-14Ni-9Al-7,5Cr 1Mo AT HIGH TEMPERATURES. Increasing the efficiency of power plant facilities, such as steam power plant, is demanding, and this drives to the development of materials capable for higher temperature operations. Ferritic steels have been widely used for boiler components due to their relatively economical and higher in thermal conductivity. Nevertheless, the application of ferritic steels for boilers is limited only at 630 ℃ due to their weakness in creep strength and corrosion at higher temperatures. Efforts to increase the performance of ferritic steels at higher temperature applications have been intensively carried out, one of which is by developing B2-(Fe,Ni)Al precipitates in the ferrite matrix. This paper discusses the microstructural evolution in alloy of Fe-14Ni-9Al-7.5Cr-1Mo (in wt.%) strengthened by B2-(Fe,Ni)Al precipitates during heating at 800, 900, and 1000 oC for 6, 20 and 48 hours. Ageing at 800 oC for 6 hours gave rounded cuboidal B2-(Fe,Ni)Al precipitates dispersed homogeneously in the ferrite matrix and changed to rounded shape at longer ageing times. At 1000 oC, however, the precipitates had rounded shape at all times of ageing. Coarsening of precipitates occurred during ageing at 800 and 1000 oC with the rate constants of 5,7 x 104 nm3/h and 2,1 x106 nm3/h respectively, which is considered relatively low. The highest hardness value of 520 HVN was observed for the sample aged at 1000 °C for 6 hours.
ANALISIS KOMPOSISI UNSUR, DENSITAS, MAKROSTRUKTUR, DAN FASA PADUAN U-6Zr-xNb PASCA UJI KOROSI Masrukan Masrukan; Deni Mustika; Deninta Andara Perdana; Jumaeri Jumaeri
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar

Abstract

ANALYSIS OF ELEMENTAL COMPOSITION, DENSITY, MACROSTRUCTURE, AND PHASE FORMATION OF U-6Zr-xNb ALLOY POST CORROSION TEST. Research on the elemental composition, density, macrostructure, and phase formation of U-6Zr-xNb alloy have been done after corrosion test. The alloy composition analysis had been done before the corrosion test which included uranium content analysis by potentiometric titration, impurity analysis with Atomic Absorption Spectroscopy (AAS), and U, Zr and Nb content analysis with X-Ray Fluorescence (XRF). The alloy composition analysis was aimed to ensure that U-6Zr-xNb fuel meets the quality requirements of nuclear fuel specifications. The density, macrostructure and phase test were done after the corrosion test. The density test is required as one of the parameters in calculating the corrosion rate, and the microstructure and alloy phase observations are required to determine the damage or corrosion products and the layers/phases formed after the occurance of corrosion. The results of the U-6Zr-xNb alloy composition analysis with XRF and potentiometric titration show that the uranium content is close to the expected level, whereas the impurity test with AAS shows high levels of impurities including Al, Fe and Si. Based on these results, U-6Zr-xNb fuel meets the requirements because the elements of Al, Fe and Si have low neutron absorption cross-section. The sample density test results of U-6Zr, U-6Zr-1Nb, U-6Zr-4Nb, and U-6Zr-7Nb were 16.9798 g/mL, 16.6115 g/mL, 15.594 g/mL, and 15.3564 g/mL respectively. Macrostructure observation of the post-corrosion alloy shows black spots which are the result of oxidation of U (IV) to U (VI). The U-6Zr-xNb alloy experiences the greatest corrosion in mineral-free water media. The damage to the alloy surface decreases as the weight percentage of Nb in the alloy increases. The results of post-corrosion alloy characterization using XRD show that the U-6Zr and U-6Zr-1Nb samples formed an α phase, while the U-6Zr-4Nb and U-6Zr-7Nb samples formed γ phase. The protective Nb2O5 oxide layer formed, however, is minute and unlikely to be detected by XRD.
PURIFICATION OF INDONESIAN NATURAL GRAPHITE AS CANDIDATE FOR NUCLEAR FUEL MATRIX BY ACID LEACHING METHOD: CHEMICAL CHARACTERIZATION Deni Mustika; Torowati Torowati; Arbi Dimyati; Sudirman Sudirman; Adel Fisli; I Made Joni; Ratih Langenati
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6059

Abstract

PURIFICATION OF INDONESIAN NATURAL GRAPHITE AS CANDIDATE FOR NUCLEAR FUEL MATRIX BY ACID LEACHING METHOD: CHEMICAL CHARACTERIZATION. Graphite matrix in Pebble Bed Reactor (PBR) – High Temperature Gas Cooled Reactor (HTGR) has an important role as heat transfer medium, neutron moderator and structural material to protect fuel. Thus, graphite matrix must fulfill chemical and physical characteristics for PBR-HTGR fuel. Indonesia has graphite sources in several regions that can potentially be purified. This research aimed to purify Indonesian natural graphite by several variation of acids and to perform chemical characterizations. Natural graphite from flotation process was purified by several variations of acid, i. e., hydrofluoric acid (HF), sulphuric acid + nitric acid (H2SO4 + HNO3) and hydrofluoric acid + hydrochloric acid + sulphuric acid (HF + HCl + H2SO4) and subsequently followed by chemical characterizations such as purity level, ash content, and boron quivalent. The highest purity was obtained in the purification process by HF with carbon content up to 99.52%; this purity level fulfills the specification of nuclear graphite (>99%). Ash content analysis shows a value in compliance with the specification requirement, i.e., < 100 ppm, and boron equivalent value also fulfills the specification value of < 1 ppm. It can be concluded from this study that the graphite purified by acid leaching with HF can be used as fuel matrix candidate but is qualified as low quality. Futher research is required to produce high quality nuclear graphite, particularly research in the minimization of the impurity by evaporation at temperatures over 950 oC to by far lower the ash content.
PROSES ELEKTRODEPOSISI UNTUK PENGUKURAN ISOTOP 242Pu DENGAN SPEKTROMETER ALPHA Boybul Boybul; Yanlinastuti Yanlinastuti; Arif Nugroho; Rosika Kriswarini; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6071

Abstract

OPTIMIZATION OF ELECTRODEPOSITION PROCESS FOR 242Pu MEASUREMENT BY ALPHA SPECTROMETRY. In this study electrodeposition method was used to design and set up an electrodeposition device for 242Pu source preparation for measurement by alpha spectrometry. The purpose of this study was to determine the optimal electrodeposition process parameters so as to obtain 242Pu alpha sources with good spectrum properties and maximum result from source preparation by electrodeposition process. Standard samples of 242Pu with specific activity were prepared from plutonium standard aqueous solution. Experiments were conducted to determine the main parameters that have significant influence on the deposition efficiency. The electrodeposition process parameters analyzed were the effects of time, electric current and anode-cathode distance using an electrolyte solution of ammonium sulphate at pH 3.5. The results of the optimization of the electrodeposition process were optimal time of 2.5 hours, electric current of 1.4A and cathode- anode distance of 10 mm. These parameters' values are the best conditions for the deposition of 242Pu. Alpha measurement results of the electrodeposited source of 242Pu also show good spectrawith electrodeposition process reproductibility of 95.3%.
ANALISIS ISOTOP 137Cs DALAM PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI DENGAN METODE PENGENDAPAN CHLOROPLATINATE DAN KOLOM PENUKAR KATION RESIN DOWEX Arif Nugroho; Boybul Boybul; Sutri Indaryati; Iis Haryati; Rosika Kriswarini; Erlina Noerpitasari
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6055

Abstract

SEPARATION OF CESIUM IN IRRADIATED U3Si2/Al FUEL PLATE WITH A DENSITY OF 4.8 gU/cm3 USING CHLOROPLATINATE DEPOSITION METHOD AND DOWEX RESIN CATION EXCHANGE COLUMN. Dissolution of the U3Si2/Al plate was carried out in hot cell 109 using 6 M HCl and 6 M HNO3 solutions. The separation of cesium and uranium in uranyl nitrate solution was carried out using chloroplatinate deposition method and a cation exchange column. In the chloroplatinate deposition, the process of removing elements that will interfere with the deposition process was performed first, then the separation of heavy elements such as uranium or plutonium was carried out by depositing the 137Cs in the form of Cs2PtCl6. The isotope content of 137Cs in the Cs2PtCl6 deposition was measured with a γ spectrometer. In the Dowex resin cation exchange column method, the uranyl nitrate solution was pipetted 150 μL to be added to 2 mL of aquadest. To this solution was then added 20 μL of Cs carrier and 1 mL of 12 M HCl. The mixture of the solution was used as feed and put into the anion exchange column containing R-Cl resin. The effluent exiting the anion exchange column was fed into the cation exchange column containing R- NH4+ resin. The 137Cs isotope bound to R NH4+ resin in the column was then eluted using 1 M HCI. The effluent was compressed until a volume of ± 2 mL was obtained. The isotope content of 137Cs in U3Si2 / Al plate obtained as a result of separation using the chloroplatinate method for middle slices coded M has an average of 0.000557 g/g fuel and 0.000652 g/g fuel for the bottom slices coded B, while the Dowex resin cation exchange column method gave an average of 0.000747 g/g fuel for middle slices coded M, and 0.000934 g/g fuel for bottom slices coded B. The recovery of 137Cs separation by the anion exchange column method was greater than that of the Cs2PtCl6 deposition method.
LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR S. Simbolon; S.R Susilowati
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2015.21.1.2254

Abstract

LABORATORY-SCALE PRODUCTION OF ADU GELS BY EXTERNAL GELATION FOR AN INTERMEDIATE HTGR NUCLEAR. The aim of this research is to produce thousands of microsphere ADU (Ammonium Diuranate) gels by using external gelation for an intermediate HTGR (High Temperature Gas-cooled Reactor) nuclear fuel in laboratory-scale. Microsphere ADU gels were based on sol-solution which was made from a homogeneous mixture of ADUN (Acid Deficient Uranyl Nitrate) which was containing uranyl ion in high concentration, a water soluble organic compound PVA (Polyvinyl Alcohol) and THFA (Tetrahydrofurfuryl Alcohol). The simple unified home made laboratory experimental machine was developed to replace test tube experiment method which was once used due to a tiny amount of microsphere ADU gels produced. It consists of four main parts: tank filled sol-solution connecting to peristaltic pump and vibrating nozzle, preliminary gelation and gelation column. The machine has successfully converted 150 mL sol-solution into thousands of drops which produced 120 - 130 drops in each minute in steady state in ammonia gas free sector. Preliminary gelation reaction was carried out in ammonia gas sector where drops react with ammonia gas in a bat an eye followed by gelation reaction in column containing ammonia solution 7 M. In ageing process, ADU gels were collected and submerged into a vessel containing ammonia solution which was shaken for 1 hour in a shaker device. Isopropyl alcohol (90%) solution was used to wash ADU gels and a digital camera was used to measured spherical form of ADU gels. Diameters in spherical spheroid form were found between 1.8 mm until 2.2 mm. The spherical purity of ADU gels were 10% - 20% others were oblate, prolate spheroid and microsphere which have hugetiny of dimples on the surface.

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