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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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CORROSION CHARACTERISTICS STUDY ON URANIUM AND ITS ALLOYS (U-Nb, U-Zr, AND U-Zr-Nb) Masrukan; Wisnu Ari Adi; Djati Handoko; Dede Djuhana
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.2025.9125

Abstract

Uranium metal (U) and its alloys, such as U-Zr, U-Nb, and U-Zr-Nb, can be used as fuel in nuclear reactors, especially in research reactors. These fuels have various characteristics, including physical, chemical, mechanical, and corrosion properties. Among these characteristics, fuel corrosion properties play a critical role during the reactor operation, after use, and also during the storage period. Fuel damage due to corrosion is influenced by several factors, such as the concentration of oxidizing agents, the flow rate of corrosive fluids, temperature, and acidity (pH). The corrosion process in uranium metal produces hydrogen gas (H₂), which reacts with uranium to form pyrophoric uranium hydride (UH₃), and uranium oxidation produces UO₂. If the oxidation process continues for an extended period, the uranium metal will be depleted. The corrosion potential (Ecorr) and corrosion rate (CR) of uranium metal in a passive state do not depend on pH until the pH reaches 7. The corrosion potential increases once the pH exceeds 7 and continues to rise to pH 11. Uranium alloys with low Zr, Nb, or Ru content, when alloyed with Zr and Nb up to 2.5 wt.% and Ru up to 5.0 wt.%, show minimal impact on the corrosion potential (Ecorr) in the pH range of 4.0–9.0 unless there is a dissolved oxidant. U-Nb alloys with higher Nb content, exhibit greater resistance to corrosion and depleted uranium (without Nb). Under acidic conditions (pH 1.18), U-6Zr-2Nb alloys demonstrate the best corrosion resistance. In demineralized water, U-6Zr-5Nb alloys show the best corrosion resistance, while under basic conditions (pH 11.02), U-6Zr-8Nb alloys exhibit the best corrosion resistance.
PENGARUH IRADIASI GAMMA TERHADAP KARAKTERISTIK POLIMER POLIESTER TAK JENUH UNTUK APLIKASI SOLIDIFIKASI LIMBAH RADIOAKTIF Risdiyana Setiawan; Purwantiningsih Sugita; Muhammad Farid; Ratiko; Jaka Rachmadetin; Kuat Heriyanto; Yuli Purwanto; Dwi Luhur Ibnu Saputra; Andry Setiawan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (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.2024.9131

Abstract

The first effort in radioactive waste management is through radioactive waste characterization by identifying nuclides and measuring radioactivity. Characterization of radioactive waste is helpful in waste management consisting of waste containment, waste decomposition, temporary waste storage, and disposal operations. The use of polyester polymers in radioactive waste solidification is one of the important solutions in safe and sustainable waste management. These polymers are chosen for their good mechanical and chemical properties, such as resistance to degradation and long-term stability. This study aims to characterize and analyze the effect of gamma irradiation on the mechanical and chemical properties of unsaturated polyester polymers used in radioactive waste solidification. Irradiation was carried out with dose variations of 50, 100, 200, and 300 kGy. The results showed that at low to medium doses (50-200 kGy), there was an increase in compressive strength and density due to the formation of crosslinks between polymer chains. However, at higher irradiation doses (300 kGy), material degradation occurred leading to a decrease in strength. Raman spectrophotometry and SEM showed changes in chemical structure and morphology due to irradiation, especially in terms of changes in functional groups and structural density. The results of this study indicate that irradiation can strengthen polymers at certain doses, but dose limits must be considered to prevent material degradation.
IMPROVEMENT OF PLASMA ELECTROLYTIC OXIDATION (PEO) COATING ON ZIRCONIUM ALLOYS USING ADDITIVES FOR NUCLEAR FUEL CLADDING APPLICATION Juan Carlos Sihotang; Maman Kartaman Ajiriyanto; Anawati
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (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.2024.9141

Abstract

This study explores the improvement of Plasma Electrolytic Oxidation (PEO) coatings on zirconium alloys using additives for pressured water reactor (PWR) fuel cladding application. PEO is a promising surface treatment for zirconium alloys, offering improved corrosion and wear resistance, essential for applications in extreme environments, such as nuclear reactors. Various types of stable particle and dissolved additives were incorporated into the electrolyte during PEO process to modify the coating's microstructure, phase composition, and mechanical properties. The incorporation of stable particle such as Al₂O₃, CeO₂, and Y₂O₃ not only reduces surface porosity and roughness but also increases the presence of the tetragonal zirconia phase (t-ZrO₂), leading to enhanced corrosion and wear resistance. Additionally, the use of dissolved additives, including complex agents and salts, positively affected plasma characteristics and further improved coating density and durability.
ANALISIS RADIONUKLIDA 137Cs DI DALAM PELAT ELEMEN BAKAR NUKLIR U3Si2/Al DENSITAS 4,8 gU/cm3 BURN UP 60% PASCA IRADIASI MENGGUNAKAN METODE KLOROPLATINAT Arif Nugroho; Komar Sutriah; Sri Mulijani; Triyono Basuki; Ahmad Ade Rofiqi; Boybul; Erlina Noerpitasari; Rosika Kriswarini; Juan Carlos Sihotang; Agus Jamaludin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (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.2024.9142

Abstract

Analysis of 137C isotopes in post-irradiation nuclear fuel is vital in determining burn up and fuel behavior during fission. The 137Cs isotope is used as input data for burn up calculations because it has a long half-life (30.17 years) and a large fission yield (6.026%). This study uses the chloroplatinate precipitation method to separate the 137Cs isotope from U3Si2/Al nuclear fuel with a density of 4.8 gU/cm³ with a burn up of 60% after irradiation. The separation process begins with cutting and dissolving the fuel sample, followed by the addition of CsNO₃ as a carrier solution and H₂PtCl₆ reagent to form Cs₂PtCl₆ precipitate. The results of the analysis using gamma spectrometry showed that the spectrum of the Cs₂PtCl₆ precipitate had the highest peak area of ​​67.6298 cps compared to the brown residue of 0.9165 cps and the supernatant of 0.2450 ​​cps. In addition, the results of this study showed an increase in the separation recovery value from 70% to 96.39%, which indicates the success of the separation of the 137Cs isotope from the U3Si2/Al fuel solution with a density of 4.8 gU/cm3 using the chloropaltinate precipitation method. The results of elemental mapping of the Cs₂PtCl₆ precipitate using SEM (scanning electron microscope) showed an even distribution of platinum (Pt), cesium (Cs), and chlorine (Cl) components, which shows the efficiency of this precipitation method. In addition, EDS (energy dispersive X-ray spectroscopy) analysis shows the elemental composition of the Cs₂PtCl₆ deposits. The three main elements identified are Pt, Cl, and Cs. The results of this study can provide an essential contribution to evaluating the performance of U3Si2/Al fuel with a density of 4.8 gU/cm3 after irradiation. They can be used for further research on Cs isotopes.
FORMULASI INHIBITOR KOROSI BERBASIS ZINK–FOSFAT DAN MIKO–DEA PADA PIPA PENDINGIN SEKUNDER REAKTOR SERBA GUNA GERRIT AUGUSTINUS SIWABESSY Rosika Kriswarini; Andon Insani; Mohammad Khotib; Komar Sutriah; Arif Nugroho; Juan Carlos Sihotang; Ade Saputra
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (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.2024.9143

Abstract

The material for the secondary cooling pipes of RSG-GAS is low carbon steel. The pipes carry fluids that have the potential to cause corrosion. To reduce the corrosion rate, both inorganic and organic inhibitors are added. The inhibitors used are a mixture of zinc-phosphate as an inorganic inhibitor and miko-DEA as an organic inhibitor in a 0.1% NaCl medium. Zinc-phosphate inhibitors are tested at concentrations of 25, 35, and 50 ppm, and miko-DEA inhibitors at concentrations of 50, 75, and 100 ppm. To study the effects on corrosion rate and the corrosion behavior of carbon steel, electrochemical methods were employed, including Open Circuit Potential (OCP) for 600 seconds, Electrochemical Impedance Spectroscopy (EIS) over a frequency range from 105 Hz to 0.01 Hz, and Potentiodynamic Plot (PDP) over a voltage range from -1 Volt to 1 Volt using a potentiostat. To support the corrosion test results, surface morphology of the samples with and without inhibitors was observed using SEM, and compounds in the zinc-phosphate and Miko-DEA solutions were identified using FTIR. The optimal formulation that resulted in the lowest corrosion rate was a mixture of 50 ppm zinc-phosphate and 75 ppm miko-DEA in a 0.1% NaCl medium. The achieved corrosion rate was 10.57 mpy.
PENGOLAHAN LIMBAH RADIOAKTIF CAIR AKTIVITAS TINGGI MELALUI VITRIFIKASI GELAS BOROSILIKAT Wati; Jaka Rachmadetin; Yuli Purwanto; Dwi Luhur Ibnu Saputra; Ratiko; Aisyah; Pungky Ayu Artiani
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (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.2024.9144

Abstract

High-level liquid radioactive waste in Indonesia is presently generated from the Mo99 radioisotope production process. Until present, the treatment of the waste has not been carried out and it is still stored in the hot cell awaiting for the establishment of the methods, technology and processing facilities. Several countries have implemented a vitrification process to process LRCAT. Several types of glass have been developed for LRCAT vitrification, namely alumina silicate glass, phosphate glass, borosilicate glass and ceramic glass. Borosilicate glass is the most ideal type of glass to use. This is because the process of making borosilicate glass is simpler and the results of vitrification of borosilicate waste-glass have chemical resistance and physical resistance as well as high waste content. The objective of this research is to determine the composition of borosilicate glass-forming materials through a vitrification process suitable for imobilation of high-level waste from 99Mo production that meets the requirements of the final waste form. Several requirements in choosing a matrix for LRCAT vitrification are: easy and practical manufacturing process, high waste content, high chemical resistance, high radiation stability, high heat stability, good physical integrity and long-term stability. The vitrification process uses borosilicate glass forming materials with two different compositions. The vitrification process was carried out at a temperature of 1150 ºC for 2.5 hours then cooled to room temperature. Glass characterization is carried out by leaching test and X-ray diffractometer (XRD). Waste glass with GF-1 glass forming material with various waste contents cannot be melted at a temperature of 1150 ºC. The vitrification process was unsuccessful and no melting of the waste-glass occurred. The glass forming material GF-1 cannot be used for LRCAT vitrification and no further research was carried out. This is related to its use on an industrial scale, if the melting temperature is higher than 1150 ºC, then the refractory corrosion rate is very high so the melter life is shorter. This will generate more solid radioactive waste (secondary waste). In addition, high melting temperature will produce more gas so that gas handling is more complex. Meanwhile, glass material with GF-2 glass forming material can be melted at a temperature of 1150 ºC. GF-1 waste-glass were not leached test because the vitrification process was unsuccessful and no melting of the waste-glass occurred. The leaching test results of all GF-2 glass-waste compositions that were made ranged from 0.79 x 10-4 gram.cm-2.day-1 to 2.25 x 10-4 gram.cm-2.day-1. The leach rate requirement for PNC Japan standard borosilicate glass is 2.30 x 10-5 gram.cm-2.day-1 and based on Pacific Northwest National Laboratory (PNNL) is 10-4 – 10-7 gram.cm-2.day-1. Based on the results of chemical resistance characterization (leachate test) and physical integrity, the composition of GF-2 glass-waste with a waste content of 20% by weight provides the most optimal characteristics
DETEKSI CACAT PADA OBJEK LOGAM DI INDUSTRI NUKLIR MENGGUNAKAN MACHINE LEARNING Tri Sulistiyo Hari Nugroho; Nazrul Effendy; Kusnanto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (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.2024.9148

Abstract

Objects made from metal are quite widely used in the industrial world, both nuclear and non-nuclear related. In the manufacturing process of this object, there is possibility of production defects occurring. Apart from that, metal objects can also experience defects during the operation of the object, for example due to impact, fracture, or due to corrosion or other chemical reactions. In the nuclear industry, metal objects include reactor tank, control rod, pump, pipes, and nuclear reactor fuel cladding. In order to ensure the continued optimal functionality of equipment, machinery, and/or installations constructed from metal, it is essential to implement a comprehensive maintenance programme. Such a programme must include regular inspections to identify any potential damage that may affect the integrity of the metal components. One method of detecting defects in technical objects that is currently being developed is using artificial intelligence or machine learning, either combined with visual inspection and analysis methods, X-ray, gamma ray, electromagnetic or ultrasonic results. Machine learning itself has been applied to several applications. The research carried out in this article used bibliographic analysis and was visualized using VOSviewer. This paper presents methodologies for the non-destructive analysis of metal objects, with a specific focus on visual analysis and machine learning. The integration of test data and visual analysis with machine learning enables the formulation of informed conclusions regarding the condition of metal objects, thereby facilitating decision-making processes.
The Effect of Current Density on The Surface Morphology in the Thin Coating Process of Nickel on Zr-2 Using the Electroplating Method Yusuf Gigih Wicaksono; Ridwan; Azwar Manaf
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 31 No. 2 (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.2025.9199

Abstract

   After the Fukushima-Daiichi reactor accident in 2011, one of the research and development focuses of nuclear fuel worldwide has been on coatings for Enhanced Accident Tolerant Fuels (EATF). Chromium (Cr) coatings are considered suitable due to their high oxidation resistance; however, Cr has limitations, particularly its poor diffusion on certain materials such as zirconium (Zr). Nickel coatings are therefore used as an interlayer to overcome the diffusion problem of chromium on zirconium substrates. Several surface coating methods are available, such as physical vapor deposition (PVD), chemical vapor deposition (CVD), high velocity oxygen fuel (HVOF), detonation gun (D-Gun), and electroplating. Electroplating was chosen in this study because of its high productivity, simple equipment, and low cost. The purpose of this research was to investigate the effect of current density on the surface morphology in the electroplating of Zr-2 with nickel. Electroplating is a process in which metal ions in an electrolyte solution are driven by an electric field to deposit onto a material. The electrolyte solution used in this study consisted of 200 g/L NiSO₄·6H₂O as the nickel source, 35 g/L NiCl₂·6H₂O as the activator, and 30 g/L H₃BO₃ as the pH buffer. The current densities applied were 0.015 A/cm², 0.025 A/cm², and 0.05 A/cm². After deposition, hardness tests were conducted, and the surface morphology was examined using SEM and EDS. The results showed that increasing the current density led to larger average nickel grain sizes, namely 4.68 μm, 6.19 μm, and 6.84 μm, as well as larger pore areas on the surface, namely 6.5 μm², 20.85 μm², and 27.98 μm². Micro-Vickers hardness tests indicated that higher current density increased hardness values, measured at 163.48 HV, 178 HV, and 234.25 HV, respectively. Cross-sectional SEM analysis revealed that the coating produced at a current density of 0.015 A/cm² showed better quality compared to higher current densities. This was evidenced by smaller pore areas, thinner coating thickness (7.44 μm compared to 19.17 μm and 8.42 μm), and the absence of defects at the coating–substrate interface, which were observed at 0.025 and 0.05 A/cm². To achieve thickness values closer to calculations, uniform spacing between the cathode and anode as well as the use of a fresh electrolyte solution are recommended. The use of a nickel interlayer can be a promising option to improve the surface performance of Zr-2. 
Ligand Interaction in Thorium Adsorption by Pacitan Bentonite Roza Indra Laksmana; Mohammad Khotib; Zainal Alim Mas'ud; June Mellawati; Afiq Azfar Pratama; Suci Indryati; Dwi Luhur Ibnu Saputra
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 32 No. 1 (2026): 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.2026.9210

Abstract

Many studies have investigated the use of bentonite as an adsorbent; however, questions remain regarding the effects of sulfate and phosphate on thorium (Th) adsorption. Sulfate may reduce adsorption through the formation of Th-sulfate complexes and competition for adsorption sites between SO2-and UO2+, whereas phosphate is an important component of subsurface systems and plays a key role in controlling Thmobility. This study investigated Thadsorption by natural and acid-activated Pacitan bentonite as adsorbents for radioactive waste treatment. Because of its large surface area and porosity, bentonite has a high adsorption capacity, which can be further enhanced by acid activation. The adsorbents were characterized by FTIR and XRD to evaluate structuralchanges after activation. Batch experiments were conducted to examine the effects of pH, contact time, temperature, and complexing ligands, including sulfate, carbonate, and phosphate, on Thadsorption. The results showed that natural bentonite exhibited high Thadsorption efficiency from the beginning of the experiment, whereas acid-activated bentonite showed a gradual increase in adsorption efficiency over time. Thcomplexation with sulfate decreased adsorption efficiency, while phosphate improved Thstability in activated bentonite. These findings indicate that Pacitan bentonite, in both natural and acid-activated forms, is a promising adsorbent for Th-containing radioactive waste, provided that the adsorption conditions are properly optimized.
Uranium Adsorption Behavior on Pacitan Bentonite Roza Indra Laksmana; Mohammad Khotib; Zainal Alim Mas'ud; June Mellawati; Afiq Azfar Pratama; Suci Indryati; Dwi Luhur Ibnu Saputra
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 31 No. 2 (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.2025.9213

Abstract

Uranium adsorption by Pacitan bentonite was investigated to evaluate its capacity for capturing uranium ions under various conditions. Before being used as an adsorbent, Pacitan bentonite was characterized in terms of pore structure, surface area, surface morphology, adsorption capacity, and chemical composition. This study aims to analyze the factors that influencing uranium adsorption on Pacitan bentonite, including contact time, pH, temperature, and the presence of ligands in solution. Both natural and acid-activated Pacitan bentonite were examined to compare their adsorption performance. The results showed that Pacitan bentonite effectively adsorbs uranium, with the highest adsorption occurring under acidic to neutral pH conditions and decreasing at higher pH levels. Acid activation reduced the uranium adsorption capacity due to structural and surface property changes. In addition, the addition of ligands such as NaCl, Na2SO4, and Na2HPO4 influenced uranium adsorption, with Na2SO4 having the most pronounced effect compared to NaCl and Na2HPO4. Overall, the results indicate that Pacitan bentonite has strong potential as a uranium adsorbent material, especially under suitable conditions to maintain ion exchange and surface adsorption. These findings provide important insights for developing adsorbent materials for nuclear waste treatment and remediation of uranium-contaminated environments. 

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