Article Per Year (5 Year)
p-Index From 2021 - 2026
0.92
P-Index
This Author published in this journals
All Journal Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Dwi Luhur Ibnu Saputra
Unknown Affiliation

Published : 6 Documents Claim Missing Document
Claim Missing Document
Check
Articles

Found 6 Documents
Search

 1 
ADSORPSI URANIUM TERLARUT OLEH KOMPOSIT AMMONIUM DIHYDROGEN PHOSPHATE/ZEOLIT Yuli Purwanto; Dwi Luhur Ibnu Saputra; Mirawaty Mirawaty; Untung Sugiharto; Wati Wati; Jaka Rachmadetin; Pungky Ayu Artiani
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 2 (2023): 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.2023.29.2.6974

Abstract

Zeolite is one of natural aluminosilicate minerals that widely studied for adsorbent for liquid waste treatment. This mineral has also been studied for treatment of waste generated from nuclear industries, including adsorption of uranium from liquid waste. The treatment of this kind of waste is critical in various steps of nuclear fuel cycle. Modification of zeolite, which can be carried out physically or chemically, was reported can increase the adsorption performance of the zeolite. In this research, the capability of zeolite was increased by adding Ammonium Dihydrogen Phosphate (ADP) to form an ADP/zeolite composite for uranium adsorption. The objective of this study is to determine the adsorption performance of ADP/zeolit composite for aqueous uranium. Natural zeolite from Lampung was used as starting material. The composite material is prepared by mixing Ammonium Dihydrogen Phosphate and zeolite in a boiling flask. The resulting composite was characterize using XRD and used for adsorption of aqueous uranium by batch method. The results show that the predominant mineral in the natural zeolite was clinoptilolite. Modification of the zeolite by ADP was succeed as was shown in the XRD pattern. The experimental results are shown by the kinetics of uranium adsorption by the ADP/zeolite composite which takes place quickly and reaches equilibrium in less than 16 minutes. The adsorption kinetics of uranium onto the ADP/zeolit composite was found to follow pseudo-second order. The optimum pH for the adsorption was 7. The uranium adsorption capacity of ADP/zeolit composite was higher than that of natural zeolite. The highest uranium adsorption was obtained for the composite with composition ratio of ADP and zeolite was 1:1.
ADSORPSI URANIUM MENGGUNAKAN Na DAN Zr – MONTMORILLONITE Dwi Luhur Ibnu Saputra; Yuli Purwanto; Risdiyana Setiawan; Moh. Cecep Cepi Hikmat; Nayandra Dias Fadilla; Pratiwi Ayuningtyas; Malik Musthofa; Mirawaty Mirawaty; Jaka Rachmadetin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 2 (2023): 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.2023.29.2.6979

Abstract

The montmorillonite can be used as an alternative for adsorption treatment of uranium waste. Cations in the montmorillonite interlayer can be used as ion exchange in uranium. Uranium liquid waste can be generated from uranium extraction process for molybdenum-99 radioisotope extraction and also in the yellow cake dissolution process. The purpose of this study is to determine the adsorption of uranium using Na montmorillonite and Zr montmorillonite. Modification of Na-montmorillonite was carried out using zirconium (IV) oxychloride octahydrate (ZrOCl2.8H2O). The Na and Zr-montmorillonite were characterized using XRD and XRF analysis. Parameters of uranium adsorption were determined by batch experiments with contact time variations of 10, 30, 60, 120, 180, 300, 360, 420 and 480 minutes and pH variations of 3, 5, 7 and 9. Kinetics of uranium adsorption were studied using pseudo first-order and pseudo second-order adsorption. The results show that the optimum pH Na-montmorillonite was pH 5 of 93,68% and Zr-montmorillonite was pH 7 of 96,52 %. The ratio of Kd values in Na and Zr-montmorillonite were 23.42 x 103 and 48.26 x 103 mL/g, respectively. The results of the first-order and second-order pseudo-plot curves show that Na and Zr-montmorillonite were in the second-order plot and constant (k2) were 0.000354 and 0.000458 (g/mg.min), respectively. In conclusion, modification of montmorillonite using Zr increased the uranium adsorption and kinetics. Therefore, this Zr-montmorillonite can be used as an alternative adsorbent for treatment of uranium liquid waste.
PENGARUH KONDISI PENYIMPANAN TERHADAP KETAHANAN KOROSI DRUM BAJA KARBON WADAH LIMBAH RADIOAKTIF Dwi Luhur Ibnu Saputra; Aisyah Aisyah; Risdiyana Setiawan; Pungky Ayu Artiani; Kuat Heriyanto; Jaka Rachmadetin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): 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.2023.29.1.6770

Abstract

The Radioactive Waste Management Installation (IPLR) in the Serpong Nuclear Area is Indonesia’s only radioactive waste management facility. This facility has treated various types of waste using carbon steel containers, including low and medium-contamination solid waste. Radioactive waste containers must have good corrosion resistance so that the waste container can survive a long storage period. Research on the corrosion of carbon steel radioactive waste containers has been carried out. This study aims to obtain data on the corrosion rate of welded materials radioactive waste container and their parent metals under disposal conditions. In the disposal facility, there is a potential for water seeps in so that the interaction with the waste container may occurred. This condition will increase the corrosion rate. In this study measured the corrosion rate in various corrosion media, namely SP4 water, demineralization water, cemented solution, and bentonite solution. The results showed that the corrosion rate of the welded container was higher than the corrosion rate of its parent metal in various corrosion media. The highest corrosion rate is SP4 water media, which is 7.86 and 6.78 times compared to demineralization water. Meanwhile, the corrosion rate in demineralization water media, bentonite water, and cement water is not significantly different. The corrosion rate is an essential parameter of radioactive waste containers because waste containers with high corrosion resistance will be able to confine radionuclides for a long period time.
KARAKTERISASI RADIONUKLIDA PADA BAHAN BAKAR BEKAS DARI EXPERIMENTAL PEBBLE BED REACTOR Aisyah Aisyah; Mirawaty Mirawaty; Dwi Luhur Ibnu Saputra; Risdiyana Setiawan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 1 (2019): 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.2019.25.1.5024

Abstract

Arbeitsgemeinschaft Versuchsreaktor (AVR) is a type of High Temperature Gas Cooled Reactor (HTGR) that has a TRISO coated pebble fuel similar to TRISO fuel for the Experimental Power Reactor (Reaktor Daya Experimental, RDE), which is planned to be built in Indonesia. Therefore, the radionuclides characterization study in the AVR spent fuel can give contribution to the study of RDE spent fuel. One of the important aspects in the operation of nuclear reactor is the management of the generated spent fuel. The management of the AVR spent fuel was carried out by storing the spent fuel in dry cask for a long time. In designing the safety of spent fuel storage system, one of the needed important studies is the characterization of radionuclides contained in the spent fuel. In this study, characterization of radionuclides contained in the spent fuel has been performed by using ORIGEN 2.1 computer program based on the data of AVR reactor operation. The objective of the research is to analyze the safety of a long-term storage of pebble spent fuels in dry cask. The results show that up to 100 years storage period, one pebble spent fuel has a total activity concentration of 4.03x1010 Bq/g of activated products, actinides and its daughter and fission product radionuclides. In a dry cask containing 1900 pebbles, the total activity concentration of the spent fuel is 7.66x1013 Bq/g. Some broken TRISO layer in some spent fuel pebbles in dry cask is evident, so it is possible that some amount of fission product radionuclides is released from the spent fuel in the dry cask such as 85Kr, 135Xe, and 131I in the form of gas, and metallic radionuclides of 137Cs,106Ru, 110mAg and 107Pd.
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.
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
Co-Authors Aisyah Aisyah Aisyah Andry Setiawan Jaka Rachmadetin Kuat Heriyanto Malik Musthofa Mirawaty Mirawaty Moh. Cecep Cepi Hikmat Muhammad Farid Nayandra Dias Fadilla Pratiwi Ayuningtyas Pungky Ayu Artiani Purwantiningsih Sugita Ratiko Risdiyana Setiawan Untung Sugiharto Wati Wati Wati Yuli Purwanto