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All Journal Jurnal Forum Nuklir Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Risdiyana Setiawan
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Materials Science & Nanotechnology

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BIOETHANOL PRODUCTION FROM COCONUT HUSK USING THE WET GAMMA IRADIATION METHOD Putra Oktavianto; Risdiyana Setiawan; Ilhami Ariyanti; Muhammad Fadhil Jamil
Jurnal Forum Nuklir JURNAL FORUM NUKLIR VOL 14 NO 2 NOVEMBER 2020
Publisher : BATAN

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/jfn.2020.14.2.5908

Abstract

BIOETHANOL PRODUCTION FROM COCONUT HUSK USING the WET GAMMA IRRADIATION METHOD. The use of coconut husk has only been used as a material for making handicrafts such as ropes, brooms, mats, and others or just burned. The combustion of coconut husk can cause air pollution. In fact, coconut husk can be used as a raw material for bioethanol production so that the beneficial value of coconut husk will also increase. One way of bioethanol production from coconut husk is by irradiating the coconut husk. The coconut husk irradiation technique to be carried out in this study is the wet irradiation technique. Wet irradiation is carried out to accelerate the process of bioethanol production because at the time of irradiation, cellulose has been hydrolyzed and glucose has been formed so that it is more efficient in time and use of the material so that the cellulose hydrolysis process is not necessary. The coconut husk samples were wet because they were mixed with 4% NaOH and were irradiated using a gamma irradiator from STTN-BATAN Yogyakarta with a dose of 30 kGy and 50 kGy and 0 kGy (or without irradiation). Then the sample is fermented with the fungus Saccharomyces Cerevisiae from tape yeast to form ethanol. Ethanol is purified and then analyzed for concentrations using pycnometric and refractometric methods. The result is that the highest ethanol content is without irradiation (0 kGy), this is due to the low dosage used. However, the main point in this wet method research is evidence of hydrolysis of cellulose by the formation of gluoxane after irradiated wet coconut husk, and with Fehling A and B analysis, brown deposits are seen proving that glucose has been formed.
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.
POTENSI LINGKUNGAN KAWASAN NUKLIR SERPONG DAN EVALUASI KESELAMATAN UNTUK TAPAK FASILITAS BOREHOLE DISPOSAL LIMBAH SUMBER RADIASI BEKAS Risdiyana Setiawan; Kartini Megasari; Sucipta Sucipta; Andry Setiawan
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.2023.29.1.6717

Abstract

Disposal of used radiation sources (DSRS) still faces several obstacles, among others, because DSRS generally have a long half-life, high levels of gamma-photon radiation, difficult options for returning to the country of manufacture, and activity that exceeds the limit for shallow land disposal. For this reason, it is necessary to develop a disposal system for DSRS waste with small-scale facilities that are economical, meet worker, community and environmental safety standards. The answer to this problem is the concept of borehole disposal (BHD). This study focuses on evaluating the environmental potential of the Serpong Nuclear Area (KNS) for the BHD facility site and its safety according to the IAEA standard, namely borehole disposal of disused sealed sources (BOSS). Environmental evaluation of KNS is carried out by characterization and evaluation of geological and non-geological environmental safety for BHD sites according to IAEA standards. In general, the results of the evaluation of environmental parameters against the criteria for the disposal site indicate a suitability. However, there are several parameters that need to be given a technological solution (engineered barrier) so that the site's function as a natural barrier can be fulfilled. Some of these technological solutions include waste packaging (capsules and containers), buffer and backfill materials, borehole linings (casing), basement plug and upper/cover plug installation. The results of the safety assessment using Borehole Disposal Concept (BDS) scoping tools show that the BHD facility concept at the KNS site is included in the green zone which means it is safe, with a maximum total dose value of 7.2e-9 Sv/y in 46,678 y. Using the BOSS standard system with the addition of an engineered barrier, the site location at KNS meets the criteria and has the potential for placement of the DSRS waste borehole disposal facility.
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.
Co-Authors Aisyah Aisyah Andry Setiawan Dwi Luhur Ibnu Saputra Ilhami Ariyanti Jaka Rachmadetin Kartini Megasari Kuat Heriyanto Malik Musthofa Mirawaty Mirawaty Moh. Cecep Cepi Hikmat Muhammad Fadhil Jamil Muhammad Farid Nayandra Dias Fadilla Pratiwi Ayuningtyas Pungky Ayu Artiani Purwantiningsih Sugita Putra Oktavianto Ratiko Sucipta Sucipta Yuli Purwanto