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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
Arjuna Subject : -
Articles 183 Documents
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PENENTUAN BERAT ISOTOP CESIUM DAN URANIUM DALAM PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 Yanlinastut Yanlinastut; Sutri Indaryati; Boybul Boybul; Supardjo Supardjo; Aslina Br. Ginting
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.2020.26.2.5874

Abstract

DETERMINATION OF WEIGHT OF CESIUM AND URANIUM ISOTOPES IN IRRADIATED U3Si2/Al FUEL PLATE WITH 4.8 gU/cm3 DENSITY AND 40% BURN UP. The fuel plate of U3Si2/Al with a density of 4.8 gU/cm3 was irradiated in the RSG-GAS multipurpose reactor with 40% burn- up. During irradiation, the amount of isotopes produced by fission reaction, such as cesium, uranium and other isotopes can be used for burn-up calculation to determine the fuel performance. Physicochemical analysis was performed to determine the weight of cesium and uranium isotopes in the irradiated fuel. The irradiated plate of U3Si2/Al was cut at the middle section (duplo) with a weight of M-1 = 0.065 g and M-2 = 0.075 g and dissolved in HCl 6N and HNO3 6N. The resulted was pipetted at 100 µL and then added into each of 2 mL solvent of HNO3, H2SO4 and HCl 0.1N. The separation of cesium was then carried out with the addition 1000 mg of zeolite from Lampung. Cesium isotope obtained from the separation was bound with zeolite in the solid phase while uranium was in the liquid phase. Cesium isotopes in the solid phase and in the liquid phase were measured by a spectrometer- for their radioactivity for 10000 seconds. The measurement was indicated by spectra of 134Cs at 604.7 and 795.8keV energy and 137Cs at 661.7keV. Meanwhile, the separation of uranium in the liquid phase was carried out using anion exchange column method with the addition of Dowex resin. The results of the separation were indicated by some spectra of U isotopes at various energies, 238U at 4,1943 MeV, 235U at 4,397 MeV, 236U at 4,494 MeV and 234U at 4,777 MeV. Weight calculation from the measurement data shows that the weight of 137Cs in 0.056 g of the irradiated U3Si2/Al fuel samples in HNO3, H2SO4 and HCl solvent are 4.08E-05 g; 4.53E-05 g ,and 4.53E-05 g respectively, while 134Cs weight are 4,84E-08 g; 5,30E-08 g and 5,20E-08 g respectively. Meanwhile, 137Cs isotope weight in 0.075 g fuel plate of U3Si2/Al are obtained by 5.93E-07 g; 5.90-05 g, and 5.88E-05 g, while the isotope weight of 134Cs in HNO3, H2SO4 and HCl solvent are 5,23E-07 g; 5,29E-07g and 5,008E-07g respectively. The isotopes weight of 238U, 235U, 236U and 234U in 0.056 g of fuel plale are 6.45E-03 g; 2.57E-03 g; 2.42E-06 g; and 1.90E-05 g, and the weight in 0.075 g fuel plale are 7.41E-03 g; 2.40E-03 g; 2.91E-06 g; and 2.17E-05 g. The isotopes weight of 134Cs, 137Cs, 238U, 235U, 236U and 234U in 0.056 g and 0.075 irradiated U3Si2/Al fuel plate obtained were relatively similarly. Data of isotopes weight are then used as input data for burn up calculation.
KOMPOSISI, STRUKTUR DAN DENSITAS PADUAN U-7Mo-xSi Supardjo; Agoeng Kadarjono; Jan Setiawan; Ganisa Kurniati Suryaman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6386

Abstract

The structure of U-7Mo-xSi (x = 1, 2, and 3%) alloy was studied to obtain a new uranium alloy suitable for use as a candidate for high-density uranium research reactor fuel. The U-7Mo-xSi alloy was made by melting technique using an arc furnace. The characterization of the alloy includes chemical composition, density, and structural phases to identify the phases contained in the alloy. The hardness test was also done. The density characterization results show that U content and density of the alloy decrease with increasing of Si content in the U-7Mo-xSi alloy. On the other hand, the hardness of the alloy increases with increasing of Si content. The X-ray diffraction pattern of U-7Mo-xSi alloy was observed from diffraction angle of 25° to 95°. The results indicate the presence of U-γ, U3Si2, and U3Si2Mo phases. The higher the Si content, the more U3Si2 phase is formed and the more the hardness increases due to the formation of bonds between Si and Si.
THE EFFECT OF ANNEALING AND COLD FORGING ON MICROSTRUCTURE AND HARDNESS PROPERTIES OF Al-SiC COMPOSITE : A PRELIMINARY STUDY Yudhistira Adityawardhana; Anne Zulfia; Bintang Adjiantoro; Muhammad Yunan Hasbi
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6396

Abstract

Aluminum Metal Matrix Composites (AMMCs) are among the exciting materials that have an extensive function in various applications. Reinforcement in the fabrication process of Al composites can produce superior properties such as high strength, good fracture resistance, and of course, lightweight. Therefore, many studies are interested in revealing the characteristics of Al composite materials through various methods and variations of reinforcement. This research was a preliminary study with a scope of work that includes observation on the effects of annealing and cold forging processes on microstructure morphology and hardness properties of SiC nano-ceramic reinforced of Al composites. The aluminum used in this study was a 7xxx series aluminum alloy. The fabrication process was carried out by stir-squeeze casting method. Microstructure analysis was conducted by optical microscopy and Scanning Electron Microscopy (SEM) equipped with Emission Dispersion Spectroscopy (EDS). The hardness properties of the Al-SiC composite were examined by micro-Vickers hardness testing. This paper reports that annealing process influences grain refinement and hardness properties of the Al-SiC composite. The sample experiencing cold forging shows improvement in the hardness value. Increase of hardness value by forging after annealing may be promoted due to grain compression as dislocation mechanism effect. Comprehensive research is required to discover other potentials of Al-SiC composite materials.
PENGARUH JENIS PREKURSOR DAN METODE PRENETRALISASI PADA MORFOLOGI PERMUKAAN KERNEL TERSINTER ZIRKONIA Sarjono; Gagad Rahmadi; Etty Mutiara; Mujinem; Dedi Haryadi
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6406

Abstract

Research activities have been carried out on the effect of precursor types and preneutralization methods on the surface morphology of sintered zirconia kernels. Surface morphology with minimum roughness is desired in the development of TRISO coating system. Several batches of broth solution were prepared by several routes using different types of precursors and preneutralization methods. The types of precursors used were zirconium oxynitrate and zirconium nitrate pentahydrate and the preparation of sol was carried out by precipitation-redispersion and urea decomposition. The experimental stages included sol preparation, broth preparation, gel-casting process, aging-washing-drying process, de-watering process, heat treatment of calcination and sintering, and visual examination of surface morphology. Sphericity measurements were carried out on dry spherical gels to study the level of saturation of the sol and its ability to convert instantly upon reaction with NH3 gas in the gel-casting process. The results of SEM visual observations of the surface of the sintered zirconia kernel show that the lowest surface roughness was given by the kernel from zirconium nitrate pentahydrate precursor with urea decomposition pre-neutralization method, and the roughest surface was given by the kernel from zirconium oxynitrate precursor with pre-neutralization method of precipitation-redispersion.
KARAKTERISASI TERMAL PEB U-7Mo/Al SEBAGAI BAHAN BAKAR REAKTOR RISET Sayyidatun Nisa; Ariyanti Saputri; Sutri Indaryati; Noviarty Noviarty; Supardjo; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6420

Abstract

UMo/Al-based fuel is a candidate for research reactor fuel and as an alternative to U3Si2/Al fuel. This is because the UMo/Al alloy has a higher uranium density of 16.4 g/cm3 compared to U3Si2/Al of 12.2 g/cm3. In addition, the UMo alloy has a low neutron absorption cross section and its reprocessing process is relatively easy. The U-7Mo/Al alloy is able to maintain the γ-U phase during irradiation process so that the stability of the fuel in the reactor is maintained. The U-7Mo alloy has been fabricated into a U-7Mo/Al mini fuel element plate. The purpose of this study was to determine the thermal characteristics of U-7Mo/Al as inputs to the research reactor fuel fabricator in the manufacture of UMo/Al-based fuel. Thermal characterization was carried out for uranium, molybdenum, UMo and the fuel element plate of U-7Mo/Al using Differential Scanning Calorimetry (DSC). The thermal characterization results show that the U-7Mo/Al fuel element plate is thermally stable up to 600 °C because it undergoes an endothermic thermochemical reaction at a temperature of 632.962 °C with DH = 144.318 J/g. At 656.186 °C, the Al matrix and AlMg2 cladding are fused in the U-7Mo/Al with DH = 144.318 J/g. At temperatures of 952.810 °C to 1106.050 °C, the molten Al matrix diffuses with UMo to form a metastable U(Mo,Al)X compound and heating to a temperature of 1211.760 °C forms a UAlX layer (UAl2, UAl3 and UAl4) with DH = –22,721 J/g.
STUDI KELAYAKAN YELLOWCAKE HASIL SAMPING PETROKIMIA GRESIK SEBAGAI BAHAN BAKU PROSES PEMURNIAN DAN KONVERSI DI FASILITAS PILOT CONVERSION PLANT Ade Saputra; Erilia Yusnitha; Imam Abdurrosyid; Ratih Langenati
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6424

Abstract

The Pilot Conversion Plant (PCP) facility has been successfully used for the purification and conversion of commercial yellowcake from Cogema – France. Thus, to support self-reliance of the nation, an assessment of the purification and conversion process of domestic yellowcake was carried out. One of domestic sources of yellowcake is the by-product of a process carried out at Petrokimia, Gresik. This study was carried out by comparing the characteristics of the chemical and physical properties of Petrokimia's yellowcake to those of Cogema's yellowcake which has been used at the PCP. Primary information was obtained from analysis results while secondary information was obtained from related literatures. The information obtained is also used to determine the process parameters needed by the PCP for the purification and conversion process. The results of this study show that Petrokimia's yellowcake is suitable to be used as the feed of purification and conversion process at the PCP facility. In addition, Petrokimia's yellowcake is easier to handle although requires more nitric acid than Cogema yellowcake dissolution. The parameters values of Petrokimia's yellowcake purification and conversion process that have been obtained include dissolution of yellowcake, purification of uranyl nitrate, uranyl nitrate concentrating, precipitation of uranyl nitrate into ammonium diuranate (ADU), drying of ADU powder, calcination of ADU into U3O8 powder and reduction of U3O8 powder into UO2 powder, which are no different from the parameters values for Cogema's yellowcake processing.
EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Kuat Heriyanto; Usman Sudjadi; Jaka Rachmadetin; Yuli Purwanto; Pungky Ayu Artiani; Siti Chotijah
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6426

Abstract

Development of dry casks is necessary to support national strategy for the management of spent fuels. One of the requirements for the dry cask is shielding performance for neutrons emitted by the spent fuels. The objectives of this study were to determine the emitted neutrons by the spent fuels generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used for dry cask for the spent fuels. The determination of the emitted neutrons was carried out using Origen 2.1 software while the evaluation of the neutron shielding performance was done using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor are 2.81×103 and 3.32×106 n/s for reactor core power of 15 and 30 MW, respectively. Addition of Cd improves the neutron shielding performance of SS 316L. It can be concluded that evaluation of neutron shielding performance of Cd-SS316L can be done using Origen 2.1 software for the neutron emission determination and MNCP 5 software for the neutron shielding performance by simulation. This study shows that the Cd-SS 316L alloy is worth further study to develop the dry cask design for the GAS research reactor
PENGARUH DISTRIBUSI UKURAN SERBUK UO2 PENGAYAAN 2%–5% PADA DENSITAS PELET UO2 Etty Mutiara; Odi Buana Hutapea; Winter Dewayatna; Tri Yulianto
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.2021.27.1.6333

Abstract

A study of UO2 pellet fabrication using 2%, 3%, 4% and 5% enriched UO2 powder without conditioning has been carried out. There is no published research on the relationship between UO2 powder size distribution and UO2 pellet density in the range of powder sizes studied in this work. Therefore, through this research the influence of size distribution and weight fraction of UO2 powder on various densities was studied to obtain UO2 sintered pellet with a density that meets the requirements as PWR fuel. UO2 pellets were fabricated through cold compaction with variation in pressure of 2, 3 and 4 ton/cm2 and sintering process at 1700 °C for 3 hours. Each UO2 feed powder with its size distribution and weight fraction was directly fabricated into pellets without conditioning.The powder sieving process shows that the 2% and 4% enriched UO2 powders have a monomodal powder size distribution while the 3% and 5% enriched UO2 powders are trimodal size. The green pellets density of 2%, 3%, 4% and 5% enriched UO2 powders are 54.3%TD, 53.19%TD, 52.45%TD and 52.33%TD respectively. The density of sintered pellets from 2% and 5% enriched UO2 powders are 91.13-91.70%TD and 75.33-77.57%TD. The sintered pellets of 3% and 4% enriched UO2 powders were cracked or broken so that the density could not be measured. Based on the pellet density data, it is concluded that the 2% enriched UO2 powders with monomodal size distribution that are dominated by fine powder smaller than 75µm has better compaction capability and sintering ability than other UO2 powders. This research needs to be continued to determine the proper size distribution of the UO2 powder and its weight fraction as compaction process feed in order to obtain higher sintered pellets density that meet the requirements of PWR fuel.
PENGARUH MEDIA PENDINGIN TERHADAP KARAKTERISTIK MEKANIK DAN STRUKTUR MIKRO PADA PELAT BAJA KARBON RENDAH Iskandarsyah Wicaksono; Sri Ismarwanti; Jan Setiawan; Ferry Budhi Susetyo; Syamsuir
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.2021.27.1.6187

Abstract

Low carbon steel has high ductility and is easy to shape, so it is widely used in industry and construction. However, this material has low mechanical strength and hardness, so it is easy to wear. Mechanical properties can be improved by heat treatment, namely heating at a specific temperature and holding it for some time and then cooling. Cooling media will affect the level of enhancement in its mechanical properties. Therefore, in this study, observations were done on the mechanical characteristics and microstructure of low carbon steel after heat treatment and cooling on various cooling media. The heating process was carried out at a temperature of 950 °C with a holding time of 15 minutes. The sample is then cooled with a variety of coolants (aqua dest, cooking oil, oil and open-door furnace). The results of the observations on samples of low carbon steel plates (0.0524%) shows that the sample without heat treatment has the highest stress value of 375.75 MPa and the open furnace door cooling has the lowest stress value of 200.53 MPa. The highest strain value occurred in the open-door cooling of 38.77% and the lowest strain value of 25.49% occurred in the cooling with cooking oil. Observation on the material's microstructure before and after heat treatment shows ferrite and pearlite.
KARAKTERISASI KANDUNGAN URANIUM DAN UNSUR JEJAK PELET SINTER UO2 UNTUK FORENSIK NUKLIR Torowati; Ngatijo; Rahmiati; Deni Mustika; Erilia Yusnitha; Tri Yulianto; Jan Setiawan
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.2021.27.1.6224

Abstract

Nuclear forensics is one of the important tools in nuclear security related to law enforcement. Nuclear materials and radioactive materials have potential hazards in terms of safety, health, up to threats to the nuclear security. In nuclear forensics, fingerprints are the characteristics of nuclear materials and radioactive materials which provide information on the origins of the nuclear materials, from where data of nuclear materials and radioactive materials are expected to be collected. The data can be obtained from characterization by physical and chemical testing. The physical tests include visual observations, dimensions, and phases while chemical tests cover elemental analysis of the nuclear materials and analysis of elements concentration in the nuclear materials. In this study, analysis of uranium and trace elements contents in uranium dioxide pellets was carried out in PTBBN - BATAN with the purpose to collect the fingerprint for nuclear forensics. The analysis of uranium content was conducted by potentiometric titration and the trace elements content was analyzed by atomic absorption spectrophotometric method. The results show that uranium content of the nuclear materials tested has the average results from 87% to 88%, indicating that the materials are UO2. The analysis of trace elements content was done to UO2 of varied enrichment level and the results show that the trace elements are not similar, which informs about the level of enrichment of the uranium dioxide pellets. These data can be used as fingerprint for nuclear forensics that can assist the investigators in the identification of nuclear forensics when misuse of nuclear materials occurs.

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