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All Journal Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Mirawaty Mirawaty
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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.
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.
ANALISIS KRITIKALITAS BAHAN BAKAR NUKLIR BEKAS REAKTOR RSG-GAS PADA RAK BERBAHAN ALUMINIUM Pungky Ayu Artiani; Mirawaty Mirawaty; Kuat Heriyanto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): 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.2017.23.2.3280

Abstract

The use of stainless steel in nuclear spent fuels storage racks of interim storage facility potentially causes galvanic corrosion so material replacement of spent fuels storage racks needs to be considered. Galvanic corrosion is due to interaction between aluminum as in nuclear fuel of Multipurpose Research Reactor - G. A. Siwabessy (RSG-GAS reactor) and stainless steel as storage rack material for nuclear spent fuel. The purpose of this study is to assess the criticality of nuclear spent fuel storage racks using aluminum material to reduce galvanic corrosion effect. criticality guarantee is necessary to maintain the safety of nuclear spent fuel interim storage facility. Optimum criticality of aluminum storage rack is assessed by varying pitch size and calculating neutron absorption rate performed under normal condition (not an accident). Criticality (keff) calculation is performed using Monte Carlo N-Particle version 6 (MCNP6) code. The model used in this study is a 3-dimensional model where racks are filled fully with spent fuels in the storage pool. The calculations show that nuclear spent fuel storage rack stainless steel materials (keff = 0.6810) compared to aluminum (keff = 0.7709) in the size of 127 mm pitch, keff increased by 13.20%, but still within the value of safety requirement (keff ≤ 0.95). keff is affected by pitch size, where the reduction in the size of 1 mm pitch can increase the keffof 14.24%. Neutron absorption rate also affects keff, where neutron absorption rate in aluminum storage racks smaller than stainless steel. Storage racks using aluminum materials meet safety aspect to use as a storage rack in nuclear spent fuel interim storage facility because it has a keff less than 0.95 on a pitch size more than 112 mm (keff = 0.9196).
Co-Authors Aisyah Aisyah Dwi Luhur Ibnu Saputra Jaka Rachmadetin Kuat Heriyanto Malik Musthofa Moh. Cecep Cepi Hikmat Nayandra Dias Fadilla Pratiwi Ayuningtyas Pungky Ayu Artiani Risdiyana Setiawan Untung Sugiharto Wati Wati Yuli Purwanto