cover
Article Per Year (5 Year)
All
Home Page
OAI Link
Editorial Team
Contact
Reviewer
Google Scholar
Contact Name
Hilda Yunita
Contact Email
jurnal@rmpi.brin.go.id
Phone
-
Journal Mail Official
urania@brin.go.id
Editorial Address
Kawasan Sains dan Teknologi (KST) Bacharuddin Jusuf Habibie, Jl. Raya Puspiptek 60, Tangerang Selatan 15310
Location
Kota tangerang selatan,
Banten
INDONESIA
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
 3   4   5   6   7 
PEMBUATAN STANDAR CESIUM DARI LARUTAN PEB U3Si2/Al PASCA IRADIASI DENSITAS 2,96 gU/cm3 DENGAN METODE PENGENDAPAN CsClO4 Aslina Br. Ginting; Boybul Boybul; Arif Nugroho; Yanlinastuti Yanlinastuti; Noviarty Noviarty; Gatot Wurdiyanto; Hermawan Chandra
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (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.2019.25.2.5457

Abstract

A considerable amount of irradiated nuclear fuel solutions exists in the hot cell area of Radiometallurgy Instalation (IRM). These solutions were generated from sampling activities for burnup test. The solutions cannot be disposed due to the present of 137Cs isotope and other heavy elements that have a long half time, but the solutions can be used as a secondary source of 137Cs standard for analysis purposes. Post-irradiation fuel analysis is always conducted by a valid method and with a calibrated spectrometer-γ. The standard used in the calibration of spectrometer-γ energy or activity is 137Cs isotope. There has been some restriction in the use of spectrometer-γ because spectrometer-γ calibration cannot be performed routinely due to the unavailability of standard sources. The availability of 137Cs isotope for R&D is still dependent on abroad suppliers. Therefore, in this research, PTBBN and PTKMR aim to develop a secondary standard source of 137Cs isotope with 10330 ± 411 Bq activity. The standard source was obtained from a solution of 137Cs separated from U3Si2/Al irradiated fuel. The separation was done by precipitation method with CsNO3 and HClO4 as the carrier. The result of the separation is 137CsClO4 precipitate, dried and weighed for obtaining greater activity and measured with a spectrometer-γ. The dried precipitate 137CsClO4 was subsequently packed as closed standard source (shield source). This standard sourced was certified by PTKMR as secondary standard isotope of 137Cs and is ready to be used for energy calibration of spectrometer-γ.
KETAHANAN MULUR DAN OKSIDASI SIKLIK BAJA SUS 304 HASIL PELAPISAN METAL MATRIX COMPOSITE NiCr+Cr3C2 DENGAN METODA HVOF Muhammad Luthfi Naufal; Muhammad Waldi; Djoko Hadi Prajitno
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 1 (2020): 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.2020.26.1.5693

Abstract

CREEP AND CYCLIC OXIDATION RESISTANCE OF SUS 304 STEEL AFTER COATING WITH METAL MATRIX COMPOSITE NiCr+Cr3C2 BY HVOF METHOD. Deposition of 60% NiCr + 30% Cr3C2 + 10% Al2O3 and 60% NiCr + 40% Cr3C2 metal matrix composite coating has been done on SUS 304 steel surface by the High Velocity Oxygen Fuel (HVOF) Thermal Spray Coating technique. The purpose of the coating was to improve SUS 304 steel creep resistance and cyclic oxidation on boiler tubes of steam power plant application. The treatments include creep test at 550oC with a load of 0.4 and 0.6 σy (yield strength) for 30240 minutes (21 days), cylclic oxidation with various temperatures of 600, 700 dan 800oC in tubes furnace for 1 hour and cooling for 15 minute in 1cycle. Measurement was done every one cycle to observe weight change of the oxide, and themeasurement was done for 5 cycles. The recommended coating specimen for boiler tubes of steam power plant application is coating specimen of 3 three components of MMC with the lowest creeprate of 0.000148917 %/hour and cyclic oxidation resistance at 600o (which has a lowest change inoxide weight, coating thickness and oxide compounds formed on coating).
CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL R. Andika Putra Dwijayanto; Ihda Husnayani; Zuhair
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (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.2019.25.2.5525

Abstract

There are several options of nuclear fuel utilization in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilizes low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilize other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fueled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0.627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yield 192.53 curies of radioactivity. These higher radioactivity and photon emission possibly necessitate different measures in spent fuel management, if RDE were to use thorium-based fuel. Tl-208 activity, which found to be emitting potentially non-negligible strong gamma emission, magnified the requirement of proper spent fuel handling especially radiation shielding in spent fuel cask.
KETAHANAN OKSIDASI ZIRCALOY-4 YANG DIPADU DENGAN YTTRIUM PADA TEMPERATUR TINGGI Perdana Immanuel; Pradoto Ambardi; Djoko Hadi Prajitno
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.2019.25.1.4663

Abstract

Yttrium is one of the rare earth metals which is used to improve properties of metals. This study was conducted to determine the effect of Yttrium (Y) on high-temperature oxidation resistance of zirconium alloy 4 by using isothermal oxidation method at temperatures of 900 °C and 1000 °C. This study employs 3 varied yttrium concentration of 0 wt%, 0.5 wt% and 1 wt% as doping material for zircalloy-4 (Zr, Sn, Fe, and Cr). The isothermal oxidation process was conducted at temperatures of 900 °C and 1000 °C for 9 hours heating of samples as cast in a furnace. Tests were done to the samples after oxidation process, which include metallographic testing, hardness testing, X-Ray Diffraction (XRD) testing and oxide thickness measurements. Test results show that the varying content of yttrium does not significantly affect the value of hardness, but there is a tendency that greater yttrium content decreases hardness value. The addition of yttrium affects the oxidation resistance at high temperatures, which can be seen from the graph of oxide thickness as well as the morphology of the oxide surface of each sample.
PERILAKU ELEKTROKIMIA BAJA TAHAN KARAT SS 316 DALAM MEDIA NANO FLUIDA Djoko Hadi Prajitno; Jan 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.2019.25.1.5053

Abstract

Corrosion is a common problem in many engineering metals and alloys. Electrochemical method is powerful instrument to use as tool to study corrosion behavior of metals and alloys in corrosive environment. The present study is to evaluate interaction between stainless steel SS 316 and nanofluid containing nano particles of ZrO2 as an additive to demineralized water by electrochemical method. Corrosion potential, Tafel polarization, and electrochemical impedance spectroscopy (EIS) of stainless steel SS 316 were performed in both demineralized water and demineralized water containing nano particle 0.05% ZrO2 at different immersion time. Corrosion potential examination shows that stainless steel 316 is actively corroded in both demin water and nanofluid. The corrosion potential tends to increase toward positive value for longer immersion time, both in demin water and nanofluid media. The value of anodic Tafel slope of also tends to increase for longer immersion time, both in demin water and nanofluid. Tafel polarization examination shows that corrosion rate of stainless steel 316 in demin water and nanofluid is not significantly different, and corrosion rate decreases with immersion time. EIS technique shows that the impedance of stainless steel 316 in nanofluid is higher when compared to the impedance in demineralized media, resulting in lower corrosion rates in nano fluids. X-ray diffraction examination of the specimens shows that the phase present in the stainless steel 316 alloys after corrosion testing is γ austenite, both in demin water and nanofluid. Stainless steel SS 316 is very good corrosion resistant material both in demin water and nanofluid with corrosion rate under 1 MPY.
ADSORPTION OF URANIUM SIMULATION WASTE USING BENTONITE: TITANIUM DIOXIDE Kris Tri Basuki; Lutfi Aditya Hasnowo; Elza Jamayanti
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.2019.25.1.4527

Abstract

Bentonite is a clay material of high surface area that have galleries within its structure. Bentonite that is modified with TiO2 will have high adsorption capability. In this study, natural bentonite and bentonite: TiO2 were characterized with FTIR, XRD and BET instruments to determine functional group, basal spacing, and specific surface area. This study also investigates the adsorption of bentonite:TiO2 in various environmental factors, such as pH (pH 1, 3, 5, and 8), contact time (10, 20, 30, 40, 50, 60, 70, 90, and 120 min), and initial uranium concentration (20, 40, 60, 80 ppm), and their influences on adsorption capacity, and determine the kinetics equation and adsorption isotherm. Based on FTIR analysis, a decrease in the band of O-H bond from water molecule was observed, which indicates the presence of TiO2 in bentonite interlayer structure. The XRD characterization of bentonite: TiO2 does not show diffraction peak in 001 plane. This is due to delamination of bentonite interlayer structure. Delamination is caused by the presence of TiO2 in large quantities, thus damaging the bentonite interlayer structure into irregular sheets. Bentonite as sheets will cause the basal spacing to increase and it is anticipated that XRD will find it difficult in detecting the 001 plane at a low 2 theta angle. The surface area of bentonite: TiO2 has increased by 12.04 m2/g. The maximum adsorption capacity of U(VI) took place at pH 5.0 for 70 minutes contact time and uranium concentration of 60 ppm. In this study, the adsorption kinetic and adsorption isotherm are pseudo second-order kinetic and Langmuir isotherm. The kinetic constant and maximum adsorption capacity of bentonite: TiO2 are 0.075 g/mg.min and 5.848 mg/g respectively.
GAMMA DOSE RATE ANALYSIS IN BIOLOGICAL SHIELDING OF HTGR-10 MWth PEBBLE-BED REACTOR Hery Adrial; Amir Hamzah; Entin Hartini
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.2019.25.1.4923

Abstract

HTGR-10 MWth is a high-temperature gas-cooled reactor. The fuel and moderator are pebble shaped with a radius of 3 cm. One fuel pebble consists of thousands of UO2 kernels with a density of 10.4 gram/cc and the enrichment rate of 17%. The core of HTGR-10 MWth is the center of origin of neutrons and gamma radiation resulting from the interaction of neutrons with pebble fuel, moderator and biological shield. The various types of radiations generated from such nuclear reactions should be monitored to ensure the safety of radiation workers. This research was conducted using MCNP-6 Program package with the aim to calculate and analyze gamma radiation dose in biological shield of HTGR-10 MWth. In this study, the biological shield is divided into 10 equal segments. The first step of the research is to benchmark the created program against the critical height of HTR-10. The results of the benchmarking show an error rate of ± 1.1327%, while the critical core height of HTGR 10 MWth for the ratio of pebble fuel and pebble moderator (F:M) of 52:48 occurs at a height of 134 cm. The rate of gamma dose at the core is 3.0052E+05 mSv/hr. On the biological shield made of regular concrete with a density of 2.3 grams/cc, the rate of gamma dose decreases according to an equation y = 0.0042 e-0.03x. Referring to Perka Bapeten no 4 of 2013, the safe limits for workers and radiation protection officers will be achieved if the minimum thickness of biological shield is 115 cm with gamma dose rate of 0 mSv/hour.
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.
SENSITIVITY ANALYSIS ON THERMOHYDRAULIC CODE FOR MODIFIED PLATE-FUELED 2 MW TRIGA Endiah Puji Hastuti; Surip Widodo
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.2019.25.1.4961

Abstract

The plan to modify TRIGA 2000 Bandung from using regular TRIGA fuel to plate-type fuel should be supported by the use of appropriate computer codes. This research proposes three codes to design reactor thermohydraulic at transient condition. Analysis has been performed to identify code sensitivity using the same input and correlation. The codes used were COOLOD-N2, Heathyd, and PARET-ANL. The input was obtained from preliminary analysis of a flow rate calculation of 70 kg/s and a nominal power of 2 MW. The comparison of these three codes did not consider uncertainty factor for neutronic and technical aspects. The sensitivity analysis on thermohydraulic codes used to calculate heat transfer in the fuel plate of TRIGA reactor at steady state condition indicates similar temperature trend lines for the coolant, plate, and fuel meat. Temperature calculation results obtained from COOLOD-N2, Heathyd and PARET ANL give consistent sensitivity with the differences of coolant temperature from 2.83% to 12.5%; cladding temperature from 2.14% to 31.30%; and fuel meat temperature from 6.63% to 18.64%. The margins of flow instability were 5.03; 5.68 and 4.21, respectively for COOLOD-N2, Heathyd, and PARET-ANL. These values show that flow instability has not yet occurred. The results of the analysis show that the use of those three codes for steady state condition using the same input, in which uncertainty factor is neglected, give similar trend for coolant, cladding, and fuel meat temperature. As the modelling in each code is different, the values obtained are not exactly the same.
PENGARUH METODE PREPARASI BROTH DAN JENIS PREKURSOR ZIRKONIUM PADA KARAKTERISTIK KERNEL YTTRIA-STABILIZED ZIRCONIA (YSZ) Sarjono Sarjono; Etty Mutiara; Erilia Yusnitha; Winter Dewayatna
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 1 (2020): 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.2020.26.1.5743

Abstract

THE INFLUENCE OF SOL PREPARATION METHOD AND TYPE OF ZIRCONIUM PRECURSOR ON THE CHARACTERISTIC OF YTTRIA-STABILIZED ZIRCONIA (YSZ) KERNEL. Fabrication of yttria-stabilized zirconia (YSZ) kernel by external gelation method has been done. Urea was used in the broth preparation to serve as: 1) as homogenous precipitation agent to produce sol when added at decomposing temperature of urea and 2) as pore promoter when added at below decomposing temperatures of urea. The research used urea in varied mol urea/metal ratio to obtain an the optimum value based on the resulted morphological properties of the sintered kernel. The zirconium precursor used was zirconium oxynitrate hydrate. Characterization of the kernel products include examination with SEM to study the morphological properties of the dried and sintered kernels, measurement of diameter and density of the sintered kernel, and XRD examination to determine the phase of the resulted solid solution of zirconium-yttrium oxide. Based on themorphological examination, the optimized addition of urea according to its function: 1) as precipitation agent in the sol preparation, mol urea/metal ratio of 1.0 at decomposing temperature of 80oC and 2) as porosity generation in the gel microsphere, mol urea/metal ratio of 1.5 at addition temperature of 50oC. The XRD examination shows that a mixture of 16% mol yttrium nitrate dan 84% mol zirkonium nitrate has resulted in solid solution of Y0,15Zr0,85O1,93 or 92ZrO2.8Y2O3 kernels with cubic phase. Increase in precursors concentration in broth has resulted in increase in the diameter of the sintered YSZ kernel, where kernels with average diameter of 0,510 mm was obtained from 0,855 mol/L precursors in broth. Kernels with desired diameter can be prepared by tuning of precursors concentration in broth.

Page 5 of 19 | Total Record : 183

 3   4   5   6   7 

Filter by Year

2015 2026


Reset
Filter By Issues
All Issue Vol. 32 No. 1 (2026): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 31 No. 2 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 31 No. 1 (2025): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 2 (2024): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 30 No. 1 (2024): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 2 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 29 No. 1 (2023): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 3 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 2 (2022): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 1 (2022): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 1 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 1 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 2 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 3 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 21 No. 3 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 21 No. 2 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 21 No. 1 (2015): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir More Issue