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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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PERBANDINGAN DENSITAS PELET UO2 HASIL PELETISASI MENGGUNAKAN SERBUK DAN MIKROSPIR Etty Mutiara; Meniek Rachmawati; Masrukan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): 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.2016.22.1.2743

Abstract

A pelletization process UO2 fuel has been developed using UO2 microsphere as a substitute of UO2 powder. Microspheres are spherical, free flowing and porous with certain hardness (soft particle). The benefit of using microsphere in pelletization process is dust free in compaction and more effective in packing so the granulation process and solid lubricants are not required. It is hypothesized that the use of UO2 microsphere in the pelletization process will provide higher sintered pellet density than UO2 powder at the same pelletization process parameters. UO2 microsphere size used in this pelletization was 900 μm with crushing strength of 2.0 N/particles while the UO2 powder size between 150 and 850 μm. The pelletization processes of UO2 microsphere and UO2 powder were performed by varying the compacting pressure between 200 MPa up to 500 MPa and sintered at temperatures of 1100 °C for 6 hours in an atmosphere of hydrogen and nitrogen gas mixture. Characterizations performed on the green and sintered pellets of UO2 microsphere and UO2 powder were dimension measurements, weighing and densities measurements. The densities of green and sintered pellets of UO2 microsphere were higher than the green and sintered pellets densities of UO2 powder with corresponded compaction pressure variations. The results indicate that the density of the green pellets both compaction results UO2 powder and UO2 microsphere increased with increasing compacting pressure. microsphere UO2 pellets density crude ranged from 82.1 to 84.2%TD. At the same sintering conditions, both Compaction UO2 powder and UO2 microsphere. Compaction shows the density increases with the greater pressure compacting process. The sintered pellets densities of UO2 obtained from this research were not appropriate with density requirements of PWR fuel. It is necessary to perform advanced research related to the effective pelletization process parameters and UO2 microsphere specifications in providing the appropriate sintered pellets densities.
PELAPISAN PERMUKAAN PELET UO2 DENGAN ZIRKONIUM DIBORIDA MENGGUNAKAN METODA SPUTTERING Sungkono; Tri Mardji Atmono; Dwi Priyantoro
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): 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.2016.22.1.2744

Abstract

Development of the nuclear fuel technology was aimed to improve nuclear power plant efficiency. One of the proposed solutions was utilization nuclear fuel with the high burn up. This leads to an increase in gas fission and nuclear reactor core reactivity. The excess reactivity of the reactor core controlled by integrated fuel burnable absorber. In connection it has made a layered thin UO2 pellets with burnable absorber. The objectives of this research was to obtain the characters of zirconium diboride layer on UO2 pellets surface i.e microstructure, crystal structure and chemical composition. The results of this experiment showed that the microstructure of UO2 + 0.4% Cr2O3 pellet has equiaxial and acicular grains with a diameter of 2.44 mm, whereas the UO2 + 0.3% Nb2O5 pellets has a grain structure in the form of equiaxial and flattened rod with a diameter of 2.47 mm. The layer of zirconium diboride on the surface of UO2 + 0.4% Cr2O3 and UO2 + 0.3% Nb2O5 pellets are similar namely thin and compact with a thickness of 2.71 mm and 2.82 mm. The identification of X-ray diffraction pattern on UO2 + 0.4% Cr2O3 and UO2 + 0.3% Nb2O5 pellets to exhibit there are of UO2 phase with cubical crystal structure and phase ZrB2 phase with hexagonal crystal structure. Meanwhile, zirconium concentration on the surface layer of UO2 + 0.4% Cr2O3 and UO2 + 0.3% Nb2O5 pellets each of 1.82 mg and 1.90 mg, the presence of zirconium element show that ZrB2 layer formed on the surface of UO2 pellets.
ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI Sri Kuntjoro; Pande Made Udiyani
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): 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.2016.22.1.2745

Abstract

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term, and it will be used as the basis for calculating the distribution of radionuclides into the environment.
ENHANCING CORROSION RESISTANCE OF AISI–1010 IN RSG–GAS SECONDARY COOLING PIPES THROUGH INHIBITOR COMPOUND INTERVENTION Enung Nurlia; Rahayu Kusumastuti; Abdul Aziz; Maman Kartaman Ajiriyanto; Rosika Kriswarini; Gadang Priyotomo; Arini Nikitasari; Geni Rina Sunaryo; Setyo Budi Utomo; Sriyono; Supriyadi; Siska Prifiharni; Siti Musabikha
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2025.8674

Abstract

The secondary cooling pipes of RSG-GAS utilize AISI-1010, facilitating water circulation directed to the cooling tower after heat absorption from the primary coolant in the heat exchanger. The subsequent dissipation of heat through the cooling tower into the ambient air aims to uphold the primary coolant's temperature below 40°C. This study focuses on evaluating the impact of inhibitor compounds on the corrosion behavior of AISI 1010. Employing FTIR and GCMS techniques, the inhibitor compounds were analyzed, complemented by visual and SEM examinations for surface morphology. The corrosion rate, influenced by the inhibitor, was quantified using a potentiostat. FTIR analysis revealed a spectrum of functional groups, encompassing O-H, N-H, aromatic compounds, C=O, C-O, and phosphate. Nitrogen, oxygen, and phosphorus elements within the inhibitor exhibited binding interactions with iron in the material. GCMS identified prominent compounds like 1H-Benzotriazole (RTs 14.4469 & 14.7746) - which is a major corrosion inhibitor indicated by its high peak and area percentage, Trans-1-methyl-2-nonyl-cyclohexane (RT 6.6834) as primary constituents of the inhibitor. Visual inspection post-immersion in a 150-ppm inhibitor solution showcased a lack of corrosion products on AISI 1010, contrasting with visible corrosion on the material without an inhibitor. SEM analysis confirmed a protective layer formation. The addition of the inhibitor at 150 ppm significantly enhanced AISI 1010's corrosion resistance, evidenced by a 46.71% reduction in the corrosion rate. This underscores the efficacy of the inhibitor in mitigating corrosion effects on AISI 1010, contributing valuable insights for materials in similar cooling system applications.
PERILAKU KOROSI BAJA KARBON RENDAH PIPA PENDINGIN SEKUNDER REAKTOR DALAM INHIBITOR ZINK FOSFAT Rosika Kriswarini; Mohammad Khotib; Andon Insani; Komar Sutriah; Noviarty; Arif Nugroho; Enung Nurlia
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/Urania.2024.8675

Abstract

Carbon steel, one of the fuction in nuclear industry is used as a material for the cooling pipe structure of nuclear reactor operating systems. In use, the pipe is susceptible to corrosion due to contact with cooling water. Inhibitors of zinc phosphate solution derived from a mixture of zinc chloride and phosphoric acid with concentration variations of 25, 35, 50, 75 and 100 ppm in 0.1% NaCl are used as materials to study the effect of corrosion rates of carbon steel. Electrochemical methods are helpful for observing the corrosion behavior of carbon steel in zinc phosphate media. Methods include Open Circuit Potential (OCP) for 600 seconds, Electrochemical Impedance Spectroscopy (EIS) in the frequency range of 105 Hz to 0.01 Hz and Potentiodynamic Plot (PDP) in the voltage range of -1 Volt to 1 Volt using potentiostat. To support the results of corrosion tests, morphological observations were carried out using SEM, chemical composition of substrat using XRF and C/S Analyzer and zinc phosphate solution content using AAS, UV-Vis and FTIR. The results obtained show that at a concentration of 35 ppm, the corrosion rate gives a lower value than at other concentrations. The corrosion rate is 11.7 mpy
MODEL SIMULATION OF THE EFFECT OF ZINC PHOSPHATE INHIBITOR ON CORROSION RATE OF SECONDARY COOLING PIPE FOR PRESSURIZED WATER REACTOR USING PREDICTOR-CORRECTOR METHOD Fajar Al Afghani; Maman Kartaman Ajiriyanto; Rosika Kriswarini; Arif Nugroho; Juan Carlos Sihotang; Ely Nurlaily; Anditania Sari Dwi Putri
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/Urania.2024.8677

Abstract

Corrosion is one of the unavoidable problems in nuclear reactor cooling systems. Secondary Cooling Pipe Multi-Purpose Reactor G.A Siwabessy (MPR-GAS) based on low carbon steel is very easy to oxidize because it is an open circulation cooling system. Zinc-phosphate inhibitor (ZnPO4) is one of the inhibitors used to inhibit the corrosion reaction that occurs between the low carbon steel cooling pipe and the cooling water environment. The research has been carried out on numerical calculations to find a solution for the corrosion rate model of steel with and without an inhibitor by the Predictor-corrector method. The simulation results show that the effect of ZnPO4 inhibitor on the corrosion rate of steel is seen as a relationship between inhibitor concentration and stable time, if the inhibitor concentration increases, the stable time will also increase. The rate of corrosion that occurs (reducing the concentration of corroded product). Therefore, the recommended inhibitor concentration value is 0.9% for models in 0.1 M HCL medium. This concentration is the most appropriate concentration to slow down the corrosion rate most minimal.
PENGUKURAN BERAT RADIONUKLIDA 137Cs DALAM LARUTAN PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI DENGAN METODE KOLOM KROMATOGRAFI Boybul; Yanlinastuti; Erlina Noerpitasari; Juan Carlos Sihotang; Helmi Fauzi Rahmatullah; Arif Nugroho; Mohammad Sukron Fajrin Husein; Rohmad Sigit; Aslina Br Ginting; Susilo WIdodo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2024.8678

Abstract

The burn up value of the fuel is one of the parameters that must be known in the post-irradiation examination. The determination of burn up aims to determine the performance of U3Si2/Al fuel after being used in the reactor. One of the isotops that can be used as an indicator of burn up is 137Cs because it has a long half-life (T1/2) of 30.17 years with a fission yield of 6.26%. Therefore, in this study, a separation was carried out to determine the amount of 137Cs in the U3Si2/Al fuel solution post irradiation using the cation exchange column method using zeolit from Lampung along with its recovery rate. The solution of irradiated U3Si2/Al fuel from Top (T), Middle (M), and Bottom (B) position, was each pipetted in a duplicate manner (duplo) in an amount of 50 µL and added with 2 mL of 0.1 M HNO3. The solutions were then measured activity using a gamma spectrometer to obtain the initial weight of 137Cs radionuclide. The solution was then inserted into a chromatography column containing 700 mg of zeolit Lampung and allowed to flow under vacuum condition. The separation results showed that 137Cs was bound to the zeolit in the form of 137Cs-zeolit solid phase, while other isotops remained in the liquid phase. The solid 137Cs-zeolit and 137Cs in the supernatant were then measured with a gamma spectrometer to determine the weight of 137Cs bound to the zeolit. The measurement results showed that the average initial weight of 137Cs (before separation) were 7.49 × 10-3 µg/gFEP and dan 6.97 × 10-3 µg/gFEP for T1 and T2 from top section, 10.50 × 10-3 µg/gFEP for M1 from Middle section, and dan 8.43 × 10-3 µg/gFEP and 8.76 × 10-3 µg/gFEP for B1 and B2 from Bottom section, respectively. After separation, the average weight of 137Cs were 7.43 × 10-3 µg/gFEP and 6.82 × 10-3 µg/gFEP for T1 and T2, 10.10 × 10-3 µg/gFEP for M1, and 8.32 × 10-3 µg/gFEP and 8.41 × 10-3 µg/gFEP for B1 and B2, respectively. From the weight of 137Cs before and after separation, the average separation recovery with the cation exchange chromatography column method using zeolit was obtained at 97.748%. This recovery value indicates that zeolit Lampung is highly selective for absorbing 137Cs in the solution of irradiated U3Si2/Al fuel.
OPTIMASI PEMISAHAN SESIUM DI DALAM PELAT ELEMEN BAKAR TIPE U3Si2/Al PASCA IRADIASI MENGGUNAKAN METODE PENGENDAPAN KLOROPLATINAT Arif Nugroho; Komar Sutriah; Sri Mulijani; Triyono Basuki; Ahmad Ade Rofiqi; Noviarty; Helmi Fauzi Rahmatullah; Ganisa Kurniati Suryaman; Guswardani
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2024.8679

Abstract

Several methods for separating fission product ¹³⁷Cs from uranium isotopes (²³⁸U, ²³⁵U, ²³⁶U, and ²³⁴U) in U₃Si₂/Al fuel have been widely developed. Optimization and separation were performed using the chloroplatinate precipitation to establish a new cesium (Cs) separation method. The method developed in this research offers the potential to achieve higher-purity cesium precipitates. This study aims to develop a method of separating the ¹³⁷Cs isotope from U₃Si₂/Al fuel element plates (PEB) using chloroplatinate reagents. The procedure involves dissolving the pre-irradiation U₃Si₂/Al PEB and adding several elements such as Sm, Sr, Cs, and Nd to represent the composition of pre-irradiation PEB, simulating its original condition. The cesium precipitation process in the supernatant solution was carried out by adding HCl and NaOH to separate impurities from cesium. The precipitation was conducted by adding H₂PtCl₆ to precipitate and isolate cesium from other elements. The precipitation process was optimized by varying the temperature and precipitation time. The optimum precipitation conditions were achieved at a temperature of -3°C and a time of 10 minutes, resulting in a Cs2PtCl6 precipitate weighing 0.3361 g. XRD and SEM analyses were used to complement and verify the cesium separation results using the chloroplatinate precipitation method, showing a uniform distribution of all elements. The purity of the Cs2PtCl6 precipitate obtained was 100%, with a diffractogram pattern identical to the reference.
PENENTUAN KEMURNIAN RADIONUKLIDA DARI PRODUK GENERATOR Mo-99/Tc-99m NON-FISI MENGGUNAKAN SPEKTROMETER GAMMA Nurul Fadila; Siska Febriana; Beta Nur Pratiwi
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2024.8680

Abstract

The medical radionuclide technetium-99m (Tc-99m) is the type of radionuclide that is most widely used in diagnostic processes in the world of nuclear medicine. This is related to its characteristic of having a half-life of 6 hours and releasing a single gamma energy (E ) of 140 keV. Tc-99m is a product of the Mo-99/ Tc-99m generator and is a decay child of molybdenum-99 (Mo-99) which has a half-life of 66 hours. The Mo-99 radionuclide can be produced via a neutron activation reaction which requires the non-fission natural target molybdenum trioxide (MoO3) as a target in a nuclear reactor, so it is called a non-fission Mo-99/Tc-99m generator. The aim of this research is to determine the purity value of the Tc-99m product from the non-fission Mo-99/Tc-99m generator using the gamma spectrometer method. This is done by analyzing the energy and efficiency calibration of the gamma spectrometer using a liquid standard source (Liquid Source) of a mixture of Ba-133 and Eu-152, and a point standard source (Point Source) of Eu-152, according to the geometry for the measurements of Mo-99 breakthrough and Tc-99m products respectively for each three-generator elution process. The neutron activation process of natural molybdenum (MoO3) was carried out in the nuclear reactor G.A. Siwabessy, Indonesia. The purity of the Tc-99m product radionuclide from the non-fission Mo-99/Tc-99m generator was obtained at 99.999% for each elution while the largest Mo-99 breakthrough was 0,0018 µCi Mo-99/mCi Tc-99m and the lowest was 1,04 x 10-4 µCi Mo-99/mCi Tc-99m. The Mo-99 breakthrough results obtained were below the value of 0,15 µCi Mo-99/mCi Tc-99m set by the US Pharmacopoeia, so the resulting product of Tc-99m was good, and showed that the product could be used for the labeling process to become a radiopharmaceutical. This radionuclide purity measurement method can be used to determine quality control parameters for non-fission Mo-99/Tc-99m generator products.
ANALISIS PENGARUH GEOMETRI FUEL ASSEMBLY TERHADAP KRITIKALITAS DAN DISTRIBUSI FLUKS NEUTRON DALAM TERAS REAKTOR RSG-GAS MENGGUNAKAN OPENMC Ibnu Fathan Rastri; Saga Octadamailah; Sukarman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/urania.2024.8681

Abstract

The G. A. Siwabessy Multi-Purpose Reactor (RSG–GAS) is included in the MTR type research reactor having a straight plate-shaped fuel element fueled by uranium silicide with a density of 2.96 gU/cm3. This research aims to analyze the influence of the geometry of the straight plate model fuel element on the criticality and distribution of neutron flux in the RSG–GAS reactor core. Monte Carlo simulations with OpenMC were carried out on five types of fuel elements namely from RSG–GAS, HANARO, WWR-SM, MARIA, and BRR, with mass U235 250 grams per fuel element. The research results show that the geometric shapes of different fuel elements influence the criticality value and flux distribution in the same core. The highest criticality value was obtained for the fuel element from HANARO followed by RSG-GAS and the lowest was obtained for the fuel element from WWR-SM. The flux distribution in the fuel elements from MARIA and BRR has a better flux distribution.

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