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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
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Articles 183 Documents
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OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4.8 gU/cm3 DENSITY Juan Carlos Sihotang; Maman Kartaman Ajiriyanto; Ely Nurlaily; Junaedi; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 3 (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.2022.28.3.6755

Abstract

To investigate the performance of AlMg2 cladding in the U3Si2/Al dispersion fuel, oxide layer characterization of AlMg2 cladding of the irradiated U3Si2/Al fuel with 4.8 gU/cm3 density was conducted. The oxide layer on the surface of AlMg2 cladding is one of the changes that occur on the cladding after theU3Si2/Al fuel plate has been irradiated in the RSG-GAS reactor to a burn-up of ∼40%. The characterization and observation of the oxide layer was conducted using SEM (Scanning Electron Microscope) and Energy-dispersive X-ray spectroscopy (EDS). Samples with a size of 3x3 mm were taken from the middle of the fuel plate (middle position). After cutting, metallographic preparation includes mounting, grinding, polishing, and ultrasonic cleaning. SEM preparation was carried out by sputter coating using Au layer. The oxide layer on the AlMg2 cladding has a thickness of 10.3 μm with a uniformly distributed cracks along the oxide layer.
SIMULASI UJI TAK MERUSAK PELAT ELEMEN BAKAR PASCA IRADIASI MENGGUNAKAN METODE PENETRANT TEST Yusuf Gigih Wicaksono; Helmi Fauzi Rahmatullah; Refa Artika; Sri Ismarwanti; Rohmad Sigit
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 3 (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.2022.28.3.6760

Abstract

The post-irradiation examination aims to determine the performance of nuclear fuel during irradiation in the reactor. Non-destructive tests at the Radiometallurgical Installation include visual observations, digital X-ray radiography, and ultrasonic tests for inspection of potential defects or anomalies in the fuel plate (PEB). Limitations in the detection of surface defects necessitate other nondestructive test methods, i.e., to simulate the use of the penetrant test method on the PEB dummy. The purpose of this simulation is to determine the ability of the penetrant test method to complement the non-destructive test data that has been previously conducted as well as to review the capability of the hot cell in applying the method. The simulation was carried out using a dummy PEB with artificial defects and used water-washable visible penetrant and solvent-removable visible penetrant. The simulation stages for implementing the penetrant test include precleaning, applying the penetrant, dwell time, removing the penetrant, applying developer, inspection, and post-cleaning. The simulation results with the penetrant test method are able to detect surface defects in the form of cracks and porosity, but cannot detect blister defects. Based on the simulation results, it can be concluded that the penetrant test can be an alternative method to complement the non-destructive test using X-ray and ultrasonic radiographic methods. In addition, to accommodate the application of the penetrant test method in IRM hot cells, it is necessary to develop support facilities while still paying attention to ease of handling and safety systems. 
RADIONUKLIDA CESIUM SEBAGAI INDIKATOR PENENTUAN BURNUP BAHAN BAKAR U3Si2/Al SECARA MERUSAK Aslina Br. Ginting; Yanlinastuti; Boybul; Supardjo; Sungkono
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 3 (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.2022.28.3.6711

Abstract

Determination of burn-up using a destructive method was carried out by separation of 137Cs in the irradiated U3Si2/Al fuel element plate (FEP) with a density of 4.8 gU/cm3. Determination of burn-up by the destructive method aims to study the suitability of the burn-up obtained by the non-destructive method or using the Origen Code software. FEP U3Si2/Al with CBBJ code 251 was cut in duplo at the top, middle and bottom with a weight of Top-1= 0.049 gFEP and Top-2=0.058 gFEP, Middle-1= 0.055 gFEP and Middle-2= 0.024 gFEP, and Bottom-1= 0.056 gFEP and Bottom-2=0.075gFEP. FEP samples were dissolved in acid to obtain a fuel solution containing fission products of 134Cs, 137Cs, 235U and other isotopes. The fuel solution was pipetted 1 mL and then transferred from the hot cell to R.135 to separate the 137Cs from 235U by a cation exchange method using zeolite from Lampung. The fuel solution was pipetted 100 μL and put into the vial in duplicate, then 1000 mg Lampung zeolite was added to the cation exchange process. The separation results were 137Cs in solid phase, while uranium and other isotopes in liquid phase. The amount of activity (Bq) of the 137Cs isotope was measured with a Spectrometer- and then used for burn-up calculation. The results of the calculation of the burn-up were 26.9714%; 55,1431%; and 37.8855% for the top, middle and bottom pieces, respectively, with an average burn-up by 40%. These values are not much different from the burn-up obtained by the non-destructive method using gamma scanning, i.e., 24.4%; 52.7% ; and 37.6% for top, middle and bottom positions with an average of 38.23% and a burn-up calculated using the Origen code, which is 40%. 
UJI PASCA IRADIASI PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 PADA BURNUP 60%: PENGAMATAN VISUAL, RADIOGRAFI SINAR-X DAN ANALISIS CITRA Rohmad Sigit; Refa Artika; Helmi Fauzi Rahmatullah; Sri Ismarwanti; Aslina Br. Ginting; Supardjo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 3 (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.2022.28.3.6786

Abstract

The development of high density U3Si2/Al fuels has been carried out by increasing the fuel density from 2.96 gU/cm3 to 4.8 gU/cm3. Increasing the density of uranium has an impact on the mechanical integrity of the cladding and its geometric stability, so a post-irradiation examination is required. The post-irradiation examination carried out included visual inspection, X-ray radiographic tests, and image analysis. Visual inspection of the fuel surface was performed using a dedicated periscope in the operating area of hot cell 102 and documented using a camera through the hot cell window. X-ray radiographic test using the voltage and current parameters at 150 kV and 1500 μA, respectively. The radiographic image is further analyzed using ImageJ software. Based on visual inspections, no blisters, swelling, excessive oxide layer, or other surface defects were found. The interpretation of X-ray radiographic images clearly shows a homogeneous fuel profile, no evidence of sub-surface defects, and a burn-up profile that matches the results of the gamma scanning test. Based on non-destructive post-irradiation examination, the overall performance of the PEB U3Si2/Al density of 4.8 gU/cm3 during irradiation on the RSG-GAS core was excellent.
PEMISAHAN DAN ANALISIS 137Cs DALAM LARUTAN ELEMEN BAKAR UJI U3Si2/Al DENSITAS 4,8 gU/cm3 DENGAN BURNUP 60% MENGGUNAKAN METODE PENUKAR KATION Yanlinastuti; Arif Nugroho; Aslina Br. Ginting; Noviarty; Boybul; Iis Haryati; Agus amaludin; Erlina Noerpitasari; Rosika Kriswarini
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 28 No. 3 (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.2022.28.3.6739

Abstract

Separation and analysis of cesium isotope in U3Si2/Al fuel with a density of 4.8 gU/cm3 and 60% burnup by cation exchange using Lampung zeolite and Dowex resin has been carried out. The purpose of this research was to determine the weight of 137Cs in the fuel. The U3Si2/Al fuel pieces by the Bottom (B), Middle (M) and Top (T) sections were dissolved in duplicate with 6 M HCl and 6 M HNO3 to obtain a U3Si2/Al fuel solution. The separation of 137Cs started with the measurement 50 μL of fuel solution in 2 mL of 0.1 M HNO3 with a gamma spectrometer to determine the initial weight of 137Cs. The solution was put into a column containing Lampung zeolite and then flowed at a speed of 0.3 mL/minute. The resulting 137Cs isotope was bound to the zeolite and effluent containing other fission isotopes. Dowex resin, Cs carrier, and HCl was added into fuel solution, then put into a column containing R-Cl- resin. The effluent that left the column was put back into the column containing the R-NH4+ resin. The 137Cs isotope that was bound to R-NH4+ resin was eluted using 1 M HCI, then the 137Cs isotope was measured using a gamma spectrometer. The results of the 137Cs isotope measurement shows the weight of each sample with code B1=0.1772 μg/g; B-2=0.1635 μg/g; M-1=0.1395 μg/g; T-1=0.1230 μg/g and T-2=0.1036 μg/g with a recovery of B-1=99.54%; B-2=98.98%; M-1=98.99%; T-1=99.38% and T-2=98.98%. Meanwhile, the average weight of the 137Cs isotope separated by Dowex resin was B-1=0.1575 μg/g; B-2=0.1470 μg/g; M-1=0.1263 μg/g; T-1=0.1140 μg/g and T-2=0.0952 μg/g with a recovery of B-1=94.80%; B-2=95.35%; M1=97.94%; T-1=98.15% and T-2=98.89%. It was concluded that the method for separating 137Cs isotopes in U3Si2/Al fuel solution by using Lampung zeolite and Dowex resin did not have a significant difference. Both methods can be used to separate 137Cs in nuclear fuel. This analysis is used to complement the data to calculate the burnup value.
STUDI PENGARUH PENAMBAHAN YTTRIUM DAN PERLAKUAN PANAS BETA TERHADAP KETAHANAN HIDROGEN PADUAN ZIRCALOY-4-0,1%Mo PADA TEMPERATUR 600°C DAN 800°C Ajra Alfatar; Eddy Agus Basuki; Djoko Hadi Prajitno
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.2022.28.2.6643

Abstract

Electrical energy is a source of energy that is widely used and needed by humans. So that the need for electrical energy will increase along with the increase in the human population. Nuclear Power Plant is one solution to this problem. One of the important components of nuclear reactors, namely nuclear fuel cladding, usually uses a material in the form of a zirconium alloy, but the use of this alloy for a long time will be susceptible to embrittlement due to hydrogen penetration. A series of experiments were conducted to determine the effect of adding yttrium and beta heat treatment on hydrogen resistance and mechanical properties of Zircaloy-4-0.1%Mo-xY alloy (x=0 wt%; 0.5 wt%; 1 wt%). The addition of yttrium causes an increase in the amount of precipitate which is thought to be α-Y while increasing the alloy hardness in both as cast alloys and beta heat treatment. Beta heat treatment decreased the alloy hardness due to an increase in the number of β-Zr phases. Hydrogen resistance can be increased by adding yttrium because it acts as a β-Zr phase stabilizer. Similarly, beta heat treatment can increase hydrogen resistance with the same method, namely increasing the number of β-Zr phases so that the amount of soluble hydrogen increases. The addition of 1 wt% Y to zircaloy-4-0.1%Mo alloy with beta heat treatment showed the best hydrogen resistance with a thickness of 6.24 μm at 600°C and 545.5 μm at 800°C.
STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm3 Juan Carlos Sihotang; Maman Kartaman Ajiriyanto; Anditania Sari Dwi Putri; Ely Nurlaily; Junaedi; Aslina Br Ginting; Supardjo
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.2022.28.2.6670

Abstract

Uranium-silicide compound fuel dispersed in aluminium matrix (U3Si2-Al) have been used in a large number of research reactors around the world because of its excellent behavior under irradiation. This fuel also provides high uranium density with typical fuel loading up to 4.8 gU/cm3 to compensate for the reduced fissile amount in LEU. To improve the density of current U3Si2-Al (2.96 gU/cm3) used in Indonesian GA Siwabessy Multipurpose Research Reactor, U3Si2-Al dispersion fuel plate with density of 4.8 gU/cm3 (U235 ∼19.75%) had been irradiated in RSG GAS for 175 days at 15 MW power to burnup level of approximately 40%. The characterization was performed using SEM-EDS and optical microscope to study microstructure of the irradiatted fuel, largely the fission gas bubbles and the interaction layer between U3Si2 fuel and Al matrix. The average diameter of the bubbles with diameter from 0.06 to 0.55 μm was 0.21 μm. The interaction layer was identified as U(Al,Si)2,3 with thickness of approximately 1.5 μm. The relatively small fission gas bubbles and the interaction layer didn’t cause swelling on the fuel and the overall performance of the fuel plate was very good.
KARAKTERISASI MATERIAL INSULASI KABEL LISTRIK TEGANGAN RENDAH DARI PRODUK LOKAL PASCA IRADIASI GAMMA Antonio Gogo Hutagaol; Muhammad Ilham Bayquni
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.2022.28.2.6615

Abstract

The aim of this research is to study the effect of gamma radiation on the semi-crystalline polymer character of local product low-voltage electrical cable insulation material, related to its use in facilities with high gamma radiation. The test samples were irradiated using a Gamma Cell device with radiation doses of 25, 50, 100, 200, 400 and 800 kGy. The degree of crystallinity and compound composition of the test material were analyzed by XRD test and characterized using FTIR. Based on the XRD test, there was no significant difference observed in the diffraction pattern between the non-irradiated test material and the irradiated test material with a large variation in dose. This indicates that there is no major phase change of the crystalline compound contained in the related test sample. The degree of crystallinity of the test sample tends to decrease as the radiation dose increases. The FTIR results show that there is a fluctuating transmittance difference between test samples with different doses of gamma radiation as well as non-gamma irradiated test samples. The presence of increasingly widening spectra at 400 kGy and 800 kGy, at wavenumber 3369 – 3370 cm-1, indicates the presence of O-H stretch clusters. Further tests are needed to strengthen the identification of functional groups, determination of organic compoundsas well as the evaluation of the mechanical character of the test sample, which is also necessary to confirm the suspected disturbance in the polymer-plasticizer association.
KARAKTERISASI RADIOACTIVE FISSION WASTE (RFW) DARI PRODUKSI RADIOISOTOP LUTESIUM-177 MENGGUNAKAN PROGRAM ORIGEN2 Lena Rosmayani; Anis Rohanda; R Farzand Abdullatif
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.2022.28.2.6672

Abstract

The types and production of radioactive waste continue to increase with the development of the use of technology so that good management is needed so as not to endanger the community or the environment. Production of radioisotopes for nuclear medicine is one of the sources of radioactive waste produced by nuclear reactors. Several radiosotopes produced at the G.A Siwabessy Multipurpose Reactor (RSG–GAS) core include Molybdenum-99, Iodine-125, Iridium-192 and Lutesium-177 (177Lu). The characterization of the 177Lu production waste was carried out using the ORIGEN2 computer program with the aim of obtaining the characteristics of the RFW waste to obtain an appropriate waste treatment technique. In this study, a characterization of radioactive waste was carried out on the production of the radioisotope 177Lu from Target Lu2O3 at various CIP positions at a power of 15 MWt and 30 MWt with a long irradiation time at each power level for 8 and 12 days. In radioisotope production, the target is irradiated in the RSG-GAS core irradiation facility, the target is placed in a quartz ampoule which is then placed in a capsule in aluminum. Then the aluminum inner capsule is inserted into the outer capsule and placed into the irradiation position. After irradiation targeting and then taking 177Lu radioisotope. Sampling of radioisotopes after irradiation forms several types of waste. One of them is radioactive fission waste (RFW) as a by-product with different properties. Apart from the target, radioactive waste can also be generated from the target capsule. Target irradiation makes the target capsule made of quartz and aluminum also activated and becomes radioactive. Therefore, it is necessary to calculate the activity concentration of the waste produced by the radioisotope 177Lu. The results of the characterization of RFW waste using the ORIGEN2 computer program with flux variations at various CIP positions at a power of 15 MWt and 30 MWt with an irradiation period of 8 and 12 days obtained a total concentration of waste activity from the production of 177Lu radioisotope which has an activity concentration between 1,06x1016 – 1,24x1016 Bq/g. Therefore, the RFW radioactive waste produced by 177Lu is classified as moderate radioactive waste based on the Government Regulation on radioactive waste management and requires careful management to ensure safety.
ESTIMASI PANAS GAMMA PADA PRODUKSI RADIOISOTOP Lu-177, Ir-192, DAN Au-198 DI TERAS MOLIBDENUM RSG-GAS Dewi Nur Riskiana; Anis Rohanda; R Farzand Abdullatif
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.2022.28.2.6654

Abstract

Gamma heating is an important issue related to the irradiation safety of a reactor and its irradiated samples. Gamma heat is generated from the interaction of gamma energy with the target material. The gamma energy produced from the reactor has different gamma energy determined by the fuel type. Uranium molybdenum (UMo) is a future fuel that has several advantages over uranium oxide (UO2) and silicide (USi), one of which to increase the reactor operatingtime. For this reason, a computational study of the RSG-GAS gamma heat with UMo fuel was carried out in several radioisotope productions such as the Lu-177, Ir-192, and Au-198. This study uses the ORIGEN code to calculate the gamma energy produced for UMo fuel types. Gamma heat estimation uses the Gamset program, designed and tested to calculate gamma heat in RSG-GAS. UMo-fueled gamma heat calculation results in RSG-GAS are 4.85 W/g ~ 8.69 W/g. This result is smaller than the gamma heat of uranium silicide (USi), which is around 9.27 W/g ~ 13.3 W/g. The radioisotope Lu-177 has the highest gamma heat of 8.69 W/g, followed by Au-198 and Ir-192, which are 5.89 W/g and 7.12 W/g, respectively. The gamma heat of the three radioisotopes produced by RSG-GAS with UMo fuel does not exceed the maximum gamma heat (20 W/g) that has been determined to ensure that it is safe for the reactor and samples.

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