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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
Arjuna Subject : -
Articles 183 Documents
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PEMODELAN DINAMIK PENDINGINAN BAHAN BAKAR NUKLIR BEKAS REAKTOR RISET SECARA NATURAL KONVEKSI PADA PROTOTIPE DRY CASK STORAGE Arifin Istavara; Ratiko; Hendra Adhi Pratama; Nasruddin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 2 (2022): 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.2022.28.2.6639

Abstract

This study specifically aims to test the feasibility of dry cask storage design as a solution for storing spent nuclear fuel (SNF) in Indonesia. Due to the limited storage space in wet type storage, this research designed, experimented and simulated simultaneously. The canister design test varies the voltage from 50 V to 125 V with closed conditions to determine the response of the canister to the voltage which is proportional to the decay heat of SNF. Experiments using dry cask storage ventilation varying the voltage from 100 V to 175 V which aims to test dry cask storage against convection natural canister cooling. Theoretical calculations and simulations using software were also carried out as a comparison of the experimental results in terms of naturally convection cooling and thermal resistance. The experimental results show the response of the canister design to function properly, the greater the voltage supplied, the higher the canister temperature, 50 V to 125 V responds from 33.4°C to 56.6°C. The canister cooling test by natural convection showed good results, the dry cask ventilation opening and closing method showed a decrease in canister temperature at a voltage of 100 Volts of 16.1 °C and 14.8 °C at 125 V. The canister temperature comparison between experiments, theoretical calculations and simulations at 175 V are 44.9 °C, 49.7 °C and 65 °C respectively, as for velocity are 0.20 m/s, 0.25 m/s and 0.39 m/s respectively. The results of theoretical calculations and simulations are slightly higher than experiments, this is due to heat loss to the environment during the experiment. The simulation results obtained temperature contours and natural convection flow behavior in the air gap, indicating the dry cask storage design is functioning properly.
EFFECT OF Al, Zr AND Mo ON CORROSION RESISTANCE OF AlXCrFeNiMo AND AlXCrFeNiZr (X=1, 1.2 AND 1.4) AS FUEL CLADDING MATERIALS Teguh Firmansyah; Bonita Dilasari; Jan Setiawan; Djoko Hadi Prajitno
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6624

Abstract

The high entropy alloy of AlxCrFeNiM (with x = 1, 1,2 and 1,4; M = Mo and Zr) was successfully synthesized using powder metallurgy technique with sintering process at 1000 oC in an inert atmosphere. These alloys were designated as fuel cladding for research reactor with high U-density fuel such as U-Mo. One of the critical in-service properties of nuclear fuel cladding is its corrosion behavior. In this study, some properties of the HEA of AlxCrFeNiM such as the phases, microstructures, hardness, and corrosion resistance in 3 wt.% NaCl solution at room temperature were investigated. The results show that the phases in the HEA of AlxCrFeNiMo are FeNi, AlNi and Mo. The HEA of AlxCrFeNiZr has more complex phases compared to the AlxCrFeNiMo. The microstructure of HEA samples show fine grains with some micropores that imperfect the solidification during the sintering process. The hardness value of the HEA of AlxCrFeNiMo has a trend of decreasing as the x value increases. The opposite trend occurs to the HEA of AlxCrFeNiZr that the hardness value increases with increasing x value. The lowest hardness value is Al1.4CrFeNiMo at 262.2 HV, and the highest hardness value of Al1.4CrFeNiZr is at 756.7 HV. The corrosion rate of the AlxCrFeNiMo does not show a specific trend with increasing x value; however, the AlxCrFeNiZr shows decreasing value with increasing x value. The lowest value for the all-HEA samples is 0.20 mmpy for the Al1.4CrFeNiZr. The results of hardness and corrosion tests show that the Zr element combined with the Al element affects not only its hardness but also its corrosion resistance.
EFEK NITROCARBURIZING PADA MATERIAL LOW CARBON STEEL TERHADAP KETAHANAN KOROSI DAN KEKERASAN Fajar Al Afghani; Erwan Hermawan; Siti Shaleha
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): 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.2022.28.1.6596

Abstract

The use of steel is currently experiencing a fairly rapid growth. Low carbon steel is widely used in radiometallurgical installation, especially for structural components of conveyors, chains, gears and so on. Its use is increasingly massive so that various kinds of developments in steel materials have been carried out. One of them is to improve the quality of steel, so it can be used for a long time. One method that can be used to improve the quality of steel is nitrocarburizing process. This research focused on the effect of the nitrocarburizing process on corrosion rate. The nitrocarburizing process was carried out on several samples with different temperatures but the same gas. Increasing the temperature in the nitrocarburizing process resulted in harder specimen surface. The increase in hardness was obtained based on the temperature difference which increased from 192 HV for the sample before dinitrocarburizing to 793 HV for the 450 ºC nitrocaburizing sample. In addition, corrosion resistance increases with increasing nitrocarburizing temperature. The nitrocarburizing process causes the diffusion of C and N atoms to form a compound layer, diffusion layer, and transition layer consisting of a layer of ε-Fe2-3N, γ’- Fe4N, Fe2(N, C)1-z, making the surface layer of the material become hard and resistant to corrosion.
ANALISA DAN INTERPRETASI DATA HASIL UJI TAK MERUSAK DALAM RANGKA PENGEMBANGAN BASIS DATA POTENSI CACAT PADA PELET UO2 Refa Artika; Helmi Fauzi Rahmatullah; Yusuf Gigih Wicaksono; Sri Ismarwanti; Rohmad Sigit
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.2023.29.2.7001

Abstract

The development of nuclear fuel for both research and power reactors has been the focus of Research Center for Nuclear Fuel Cycle and Radioactive Waste Technology (PRTDBBNLR) research from fabrication, pre-irradiation, irradiation, and post-irradiation examination phases. As the only post-irradiation examination facility in Southeast Asia, Radiometallurgy Installation (IRM) is required to be ready to evaluate the performance of nuclear fuel, one of them is the short pin PWR-fuel containing UO2 pellets which is currently in the irradiation process. One of the efforts made by IRM is to develop non-destructive testing techniques, including methods, analysis, interpretation, and inventory of potential defects that may occur systematically in short pin PWR-fuel containing UO2 pellets. The aim of this research is to inventory methods and data of potential defects in UO2 pellets for a post-irradiation examination database. Characterization was performed on five samples of pre-irradiated UO2 pellets with variations in sintered/raw, good/defect conditions, and differences in uranium enrichment (3%, 4%, and 5%) using visual test methods and digital X-ray radiography. Visual examination can provide interpretation results regarding the conditions on the surface of the UO2 pellets. Digital X-ray radiography was operated at a voltage of 150 kV and a current of 1800 µA to produce a digital image then be carried out an image enhancement process using an image processor to interpret the UO2 pellets. Digital X-ray radiography method and image enhancement process adds information of UO2 pellets data. Cracks and surface holes of UO2 pellets can be mapped. The difference in uranium enrichment can be observed from the gray value graphic. Information from this interpretation will be useful as an initial database of interpretation methods to determine the condition of the UO2 pellets in the irradiated short pin PWR-fuel.
KOMPARASI PERFORMA MONITOR RADIASI GAMMA DALAM PEMANTAUAN RADIASI REAL–TIME Rokhmat Arifianto; Robi Sulaiman; Slamet; Ricard Parulian Hutabarat; Roni Cahya Ciputra; Frederikus Dian Indrastomo
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.2023.29.2.6981

Abstract

The National Research and Innovation Agency has a research facility used for research on nuclear materials that have potential radiation hazards that can endanger workers. To reduce the potential radiation hazard, radiation monitoring is carried out routinely by workers. To optimize radiation monitoring, three (3) radiation monitors were developed using 1 (one) scintillation (S) radiation detector and 2 (two) GM detectors (GM and GMT) which can later be installed and monitor radiation in real-time. Testing of the 3 radiation monitors developed was carried out in this study to determine the measurement accuracy of each radiation monitor. The tests carried out were linear regression tests to obtain a conversion model from units of counts per second to units of microsivert per hour. In addition, ANOVA testing was conducted to see if there was a significant difference between the dose rate measurements from the three radiation monitors compared to the calibrated surveymeter. Tukey HSD testing was carried out to test each of the developed radiation monitors and compared with a calibrated surveymeter. The results of linear regression testing between the surveymeter (GS) and the three radiation monitors obtained a coefficient of determination above 0.95. ANOVA testing conducted found that there was a significant difference in the average radiation dose rate measurement results from each radiation monitor. Tukey HSD test results show that only one of the radiation monitors has an average radiation dose rate measurement value that is not significantly different from the average radiation dose rate measurement value of the calibrated surveymeter. Therefore, the GMT radiation monitors that have been developed are radiation monitors that are suitable for use in radiation monitoring at nuclear materials research facilities.
VERIFIKASI PENGUKURAN RADIOAKTIVITAS 153Sm MENGGUNAKAN DOSE CALIBRATOR MULTITIPE Ligwina Dita Pertiwi; Maskur Maskur; Yanto Yanto; Nuha Fairusya; Nuri Jannati Wahyu Ekaningsih
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.2023.29.2.6838

Abstract

Measurement verification is crucial to ensure measurement results validity. In this study, we have done verification on 153 Sm radioactivity using five different types of dose calibrators dose calibrator Atomlab 100, Atomlab 300, Atomlab 400, Atomlab 500, and Capintec CRC-tR5. The parameters of verification process including accuracy, precision, linerity, Limit of Detection (LOD), and Limit of Quantification (LOQ). The dose calibrator Atomlab 300 was established as the standard instrument to examine the others type of dose calibrator accuracy. Verification results showed all the instruments have an error of less than 3%. However, the t-test results exhibited only the dose calibrator Atomlab 400 that had tvalue less than the corresponding ttable value. In contrast, Atomlab 100, Atomlab 500, and Capintec CRC tR5 produced a tvalue greater than a ttable value. In the precision test, all types of dose calibrator gave a satisfactory precision results. All five-dose calibrators demonstrated good linearity for measuring radioactivity within the range of 20-140 mCi. The LOQ values for each dose calibrator were as follows dose calibrator Atomlab 100 = 7,88 µCi, Atomlab 300 = 12,65 µCi, Atomlab 400 = 17,32 µCi, Atomlab 500 = 15,06 µCi, and Capintec CRC tR5 = 18,6 µCi. As well as the LOQ values of the dose calibrator were determined as calibrator Atomlab 100 = 2,74 µCi, Atomlab 300 = 1,14 µCi, Atomlab 400 = 8,2 µCi, Atomlab 500 = 6,79 µCi, and Capintec CRC tR5 = 7,38 µCi. These showed that five dose calibrator types offer reliable measurement validity with a 95% confidence level. Nevertheless, it is necessary to employ a correction factor for Atomlab 100, Atomlab 500, and Capintec CRC tR5.
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.
ANALISIS PERILAKU ISOTERM, TERMODINAMIKA, DAN KINETIKA ADSORPSI STRONTIUM MENGGUNAKAN TiO2 – ZEOLIT LAMPUNG M. Fiqri Sholahudin; Henny Purwaningsih; Erlina Noerpitasari; Boybul Boybul; Yanlinastuti Yanlinastuti
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.2023.29.2.6840

Abstract

After being used in the reactor, nuclear fuel produces several fission products, one of which is strontium (90Sr). Strontium isotopes can be used as raw material for making batteries, but the strontium is still mixed with fission products or other isotope so a separation process is needed. One method of separating strontium is by adsorption. In this research, the Lampung TiO2-zeolite mixture was evaluated for its performance as an adsorbent so that it can be used for strontium adsorption. Lampung Zeolite was initially activated using concentrated sulfuric acid to clean pores that still contained impurities. Then the zeolite is mixed with TiO2 using the stir casting method. The Lampung TiO2-zeolite adsorbent obtained was then characterized using Fourier Transform Infrared (FTIR), X-Ray Diffraction Spectrometer (XRD), and Scanning Electron Microscope (SEM). After that, adsorption tests were carried out using isotherm, thermodynamic and kinetic models. The formation of TiO2-zeolite which was characterized using FTIR spectroscopy was shown by the presence of a Ti-O-Si absorption peak in the Lampung zeolite after TiO2 was added. The results of analysis using XRD show that the Lampung zeolite is clinoptilolite and mordenite, and the TiO2 used has an anatase phase. The results of analysis using SEM indicate that TiO2 is distributed on the surface of the Lampung zeolite. The adsorption isotherm model for the TiO2-zeolite adsorbent fulfills the Langmuir isotherm model. Thermodynamic studies show that the adsorption process is endothermic and not spontaneous, and the kinetics of strontium adsorption follows pseudo second order.
PENGARUH ROL DINGIN TERHADAP SIFAT MEKANIK MATERIAL AL 6061-T6 Saga Octadamailah; Dwi Marta Nurjaya; Supardjo; Ganisa Kurniati Suryaman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 1 (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.2021.27.1.6339

Abstract

PENGARUH PROSES PENGEROLAN DINGIN TERHADAP SIFAT MEKANIK MATERIAL AL 6061-T6. Proses pembuatan pelat tutup dan pelat bingkai untuk pelat elemen bakar reaktor G.A. Siwabessy dilakukan dengan cara dirol dingin, sehingga dicapai ketebalan yang diinginkan. Pada penelitian sebelumnya, proses pengerolan dingin pada material AlMg2 menyebabkan perubahan sifat mekanik, dimana material menjadi lebih ulet dan kekuatan tariknya menurun. Pada bahan bakar densitas tinggi, material kelongsong Al 6061-T6 akan digunakan untuk menggantikan material kelongsong sebelumnya, yaitu AlMg2. Tujuan dari penelitian ini adalah untuk mengetahui perubahan sifat mekanik dari material Al 6061-T6 hasil pengerolan. Untuk melihat perubahan sifat mekanik yang terjadi, dilakukan uji tarik, pengamatan mikrostruktur dan struktur kristal. Hasil uji tarik menunjukkan bahwa kekuatan tarik, modulus elastis dan kekuatan luluh cenderung naik seiring dengan besarnya derajat reduksi ketebalan. Meskipun demikian, kekuatan luluh dari hasil pengujian cukup rendah. Rasio antara kekuatan luluh dan kekuatan tariknya hanya 0,7. Dari hasil pemindaian menggunakan difraksi sinar-X diketahui bahwa proses pengerolan dapat menyebabkan penghalusan ukuran kristalit. Kesimpulan dari penelitian ini secara umum adalah proses pengerolan dapat meningkatkan sifat mekanik material Al 6061-T6.

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