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CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage Pungky Ayu Artiani; Yuli Purwanto; Aisyah Aisyah; Ratiko Ratiko; Jaka Rachmadetin; Kuat Heriyanto
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 3 (2021): October (2021)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.3.6355

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff value of the proposed dry cask design for the RDNK spent fuel. The keff values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.

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): Juni, 2021
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.2.6426

EVALUATION OF NEUTRON SHIELDING PERFORMANCE OF CD-SS 316L AS A CANDIDATE ALLOY FOR DRY CASK OF RESEARCH REACTOR SPENT FUEL Development of dry casks is necessary to support the national strategy for management of spent fuels. One of the requirements for the dry cask is shielding performance for neutron emitted by the spent fuels to be stored in the dry cask. The objectives of this study are to determine the emitted neutrons by the spent fuel generated from GAS research reactor and to evaluate the neutron shielding performance of Cd-SS316L alloy as a candidate material to be used in dry cask for the spent fuels. The former was carried out using Origen 2.1 software, while the latter using MCNP5. The result shows that the emitted neutrons by a spent fuel after 5 years discharged from GAS research reactor were 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. The evaluation of neutron shielding performance of SS 316L with addition of Cd which is the candidate material for dry cask of the spent fuels from the GAS research reactor can be evaluated using Origen 2.1 software for neutron emission, while the neutron shielding performance was evaluated by the simulation using MNCP 5 software. This study shows the Cd-SS 316L alloy can be used for further study to develop the dry cask design for the GAS research reactor.Key words: Neutron shielding, cadmium, stainless steel, spent fuel.

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

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

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

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.

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

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 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

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

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

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