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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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MIKROSTRUKTUR DAN KOMPOSISI FASA PELET SINTER UO2 DENGAN PENAMBAHAN DOPAN LOGAM ZIRKONIUM Tri Yulianto; Etty Mutiara
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): 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.2018.24.1.3941

Abstract

Characterization of zirconium-doped UO2 pellet for PWR fuel has been done. The investigation aims to study the influence of zirconium doping on microstructure, phase composition and stoichiometric (O/U) ratio. Homogenous distribution of zirconium metal is expected in order to enhance UO2 pellet quality. The UO2 pellet was fabricated by compaction and sintering with varied zirconium addition of 0.3, 0.5 and 0.9 wt%. Sintering was done at 1700 °C with heating rate of 250 °C/hour for 3 hours in hydrogen medium. Characterization of the sintered pellet includes visual examination, density measurement, microstructure examination with optical microscope at 500x magnification and phase analysis by X-ray diffraction (XRD) method. Analysis results show that the addition of varied Zr of 0.3, 0.5 and 0.9 wt% did not cause defect or crack of the sintered UO2 pellet. Microstructure examination with optical microscope, however, could not visually confirm the existence of zirconium. The evaluation of XRD data was done with HighScore software by conforming peaks data of the test sample with peaks data of a reference database. The phase of UO2 pellet without zirconium addition was confirmed to be UO2 and UO3. The phase of UO2 pellet with zirconium addition, on the other hand, shows that the UO3 fraction increases linearly with the percentage of zirconium added, which is indicated by the value of stoichiometric (O/U) ratio.
KARAKTERISASI PADUAN AlMgSi UNTUK KELONGSONG BAHAN BAKAR U3Si2/Al DENSITAS URANIUM 5,2 gU/cm3 Aslina Br. Ginting; Supardjo; Yanlinastuti Yanlinastuti; Dian Anggraini; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): 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.2018.24.1.4016

Abstract

Increase in uranium density from 2.96 gU/cm3 to 5.2 gU/cm3 of U3Si2/Al fuel requires a compatible cladding. High density fuel possesses high hardness value, and the use of AlMg2 as cladding material may cause dog bone during rolling process. In addition to dog bone phenomena, swelling can also occur during the use of nuclear fuel in the reactor due to rising in fission products and burn up. For this reason, substitution of cladding material should be determined to suit the high density U3Si2/Al fuel. This research deals with the characterization of AlMgSi as a candidate for AlMg2 substitution. The characterization includes thermal analysis with DTA (Differential Thermal Analysis) dan DSC (Differential Scanning Calorimetry), hardness test with macrohardness tester, microstructure examination with SEM (Scanning Electron Microscope), and corrosion analysis by heating at 150 °C for 77 hours in an autoclave. Analysis results show that both AlMgSi and AlMg2 share stable heat compatibility with U3Si2/Al up to 650 °C. AlMgSi cladding has a hardness value of 115 HVN, higher than that of AlMg2, which is 70.1 HVN. Microstructure examination shows that AlMgSi has better morphology of interface bonding than AlMg2. Corrosion rate analysis also indicates that AlMgSi has lower corrosion rate. The results of thermal, hardness, microstructure and corrosion characterization reveal that the U3Si2/Al fuel of 5.2 gU/cm3 suits better with AlMgSi than AlMg2.
ANALISIS METALOGRAFI PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 2,96 gU/cm3 PASCA IRADIASI Maman Kartaman Ajiriyanto; Aslina Br. Ginting; Junaedi
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): 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.2018.24.1.4010

Abstract

Metallographic analysis of U3Si2/Al fuel element plate has been performed in hot cell. The purpose of metallographic analysis is to study changes in PEB U3Si2/Al microstructure and AlMg2 cladding thickness after irradiation in reactor until burn up of 56%. The fuel element plate of irradiated U3Si2/Al was cut in top, middle and bottom positions with each size around 5x5x1,37mm. Metallographic preparation starts from sample cutting using cutting machine with low speed and sample mounting, grinding and polishing in hot cell 104–105. Sample mounting was done by using resin for more than 10 hours followed by grinding with sandpapers up to grit size of 2400 and polishing with diamond paste of size 3 to 1 micron at a rotational speed of 150 rpm for 5 minutes. Microstructure observation was performed with optical microscope in hot cell 107 at 200 times magnification. Microstructure examination reveals U3Si2 particles with diverse forms and sizes, Al matrix and AlMg2 cladding were spread along the U3Si2/Al side. Microstructure observation of irradiated U3Si2/Al has not shown good result because only topography observation of U3Si2/Al meat, Al matrix and AlMg2 cladding can be done due to limited capability of the optical microscope in hot cell, where maximum magnification can be attained only at 200 times so that the phenomenon of interaction layer and small gas bubble cannot be observed. However, U3Si2/Al microstructure of 56% burnup, if compared to the microstructure of U3Si2/Al fuel element plate of 60% burnup from previous researcher, shows interaction between U3Si2 meat with Al matrix and the existence of layers with a thickness about 5 up to 20 microns. Meanwhile, the observed thickness of AlMg2 cladding is greater than 0.25 mm, which indicates that irradiation does not significantly change the thickness of AlMg2 cladding so that the overall irradiated U3Si2-Al still has good integrity and stability.
ANALISIS PROBABILISTIC FRACTURE MECHANICS PADA EVALUASI KEANDALAN BEJANA TEKAN REAKTOR SECARA 3-D Entin Hartini; Roziq Himawan; Mike Susmikanti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): 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.2018.24.1.3721

Abstract

Material integrity analysis is required for the reactor pressure vessel (RPV). The component is the pressure boundary that serves to contain radioactive materials. The presence of cracks on the walls can affect the integrity of the RPV. The objective of this study is to analyze fracture mechanics using probabilistic models for the evaluation of RPV reliability. Probabilistic models are used as an analysis approach for random character of input quantities such as the mechanical properties of materials and the physical environment. The random character of input quantities was obtained by sampling technique base on probability density function. The material used in this study was ferritic steel (SA 533). Fracture mechanics analysis was performed by Finite Element Method (FEM) with MSC MARC software. The outputs of MSC MARC, which is of J integral values, were converted into stress intensity factor (SIF) for evaluating the reliability of RPV in 3D model. It is found that the calculation resulted from the SIF by probabilistic method reaches a limit value of fracture toughness earlier than the SIF by deterministic method. The SIF generated by probabilistic method is 95,8 MPa m0.5, while the SIF generated by deterministic method is 91,8 MPa m0.5, where lower ratio value of (a/c) will produce greater value of SIF.
DESIGN OF DRY CASK STORAGE FOR SERPONG MULTI PURPOSE REACTOR SPENT NUCLEAR FUEL Dyah Sulistyani Rahayu; Yuli Purwanto; Zainus Salimin
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): 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.2018.24.1.3694

Abstract

The spent nuclear fuel (SNF) from Serpong Multipurpose Reactor, after 100 days storing in the reactor pond, is transferred to water pool interim storage for spent fuel (ISFSF). At present there are a remaining of 245 elements of SNF on the ISSF,198 element of which have been re-exported to the USA. The dry-cask storage allows the SNF, which has already been cooled in the ISSF, to lower its radiation exposure and heat decay at a very low level. Design of the dry cask storage for SNF has been done. Dual purpose of unventilated vertical dry cask was selected among other choices of metal cask, horizontal concrete modules, and modular vaults by taking into account of technical and economical advantages. The designed structure of cask consists of SNF rack canister, inner steel liner, concrete shielding of cask, and outer steel liner. To avoid bimetallic corrosion, the construction material for canister and inner steel liner follows the same material construction of fuel cladding, i.e. the alloy of AlMg2. The construction material of outer steel liner is copper to facilitate the heat transfer from the cask to the atmosphere. The total decay heat is transferred from SNF elements bundle to the atmosphere by a serial of heat transfer resistance for canister wall, inner steel liner, concrete shielding, and outer steel liner respectively. The rack canister optimum capacity of 34 fuel elements was designed by geometric similarity method based on SNF position arrangement of 7 x 6 triangular pitch array of fuel elements for prohibiting criticality by spontaneous neutron. The SNF elements are stored vertically on the rack canister. The thickness of concrete wall shielding was calculated by trial and error to give air temperature of 30 oC and radiation dose on the wall surface of outer liner of 200 mrem/h. The SNF elements bundles originate from the existing racks of wet storage, i.e. rack canister no 3, 8 and 10. The value of I0 from the rack no 3, 8 and 10 are 434.307; 446.344; and 442.375 mrem/h respectively. The total heat decay from rack canister no 3,8 and 10 are 179.640; 335.2; and 298.551 watts. The result of the trial-and-error calculation indicates that the rack canister no 3, 8 and 10 need the thickness of concrete shielding of 0.1912, 0.1954 and 0.1940 m respectively.
PENGARUH TEMPERATUR DAN KOMPOSISI Nb TERHADAP INTERDIFUSI PADUAN U-Zr-Nb Maman Kartaman Ajiriyanto; Masrukan; Fajar Al Afghani; Yanlinastuti Yanlinastuti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (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.2021.27.3.6503

Abstract

A simulation of U-Zr-Nb intermixed fuel has been carried out with aluminum plates. Initially, U-6Zr-2Nb, U-6Zr-5Nb, and U-6Zr-8Nb alloy ingots were made using an electric arc melting furnace. The ingot obtained was cut into pieces with a thickness of 2-3 mm and a diameter of about 5-6 mm, then it was covered Al plate on one side and pressed. After pressing, the alloy ingots of U-6Zr-2Nb, U-6Zr-5Nb, and U-6Zr-8Nb were heated at temperatures of 350 and 450 oC for 1 hour. The ingot was subjected to several tests, namely microstructure examination using optical microscope and Scanning Electron Microscope (SEM) equipped with Energy Dispersive Spectrometer (EDS), hardness test using microhardness tester, and phase observation using X-ray diffractometer (XRD). The results of microstructure examination show that the boundaries between the U-6Zr-2Nb, U-6Zr-5Nb, and U-6Zr-8Nb layers (interfaces) and the Al layer were clearly visible. The higher of the Nb content, the denser the interface between the two layers. For the microhardness test, we found that the hardness at the interface was difficult to measure, while the U-Zr-Nb region shows an increase in hardness for heating from 350 to 450 oC, while in the Al section there is a decrease in hardness when the heating temperature increases from 350 to 450 oC. UO3 compounds were formed in all test samples while the γ-U phase was formed in U-6Zr-2Nb samples both heated at temperatures of 350 and 450 oC.
PERILAKU TARIK DAN STRUKTUR MIKRO BAJA TAHAN KARAT AISI 304 PASCA PERLAKUAN PANAS PADA DAERAH SENSITISASI 600 – 700 °C Sungkono; Sri Ismarwanti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (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.2021.27.3.6570

Abstract

AISI 304 stainless steel is used as a structural component of the LWR type power reactors. In the manufacture of structural component, the quality of SS 304 welded joints is determined by cooling process from the austenite to sensitization temperature. The purposes of this study were to determine tensile strength, hardness and microstructure of SS 304 in the sensitization area of 600 – 700 °C. The research method was testing of the chemical composition, tensile, hardness and microstructural properties referring to ASTM. The results show that the tensile strength of SS 304 after sensitization treatment at 600 °C (30-60 minutes) and 700 °C (30-90 minutes) was relatively the same and lower than samples without heat treatment while the sample treated at 600 °C (90 minutes) was relatively the same as the sample without heat treatment. The microstructure of SS 304 consists of d-ferrite and austenite phases and chromium carbide precipitates. The quantity of precipitate at the grain boundaries decreases with increasing temperature and holding time. The hardness after sensitization treatment at 600 – 700 °C was relatively the same on the base metal, near the fracture and the fracture area. The most crack resistant was obtained on SS 304 samples with treatment at 700 °C (90 minutes).
ANALISIS KETIDAKPASTIAN PENGUKURAN KEKUATAN TARIK DAN ELONGASI SPESIMEN SS304 BERBENTUK RING Sri Ismarwanti; Jan Setiawan; Maman Kartaman Ajiriyanto; Rohmad Sigit
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (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.2021.27.3.6567

Abstract

Uncertainty is a measure of the reliability of a measurement. One of the widely used uncertainty analysis approaches is the Guide to the expression of uncertainty in measurement (GUM). The steps for analyzing the uncertainty of a measurement have been described in detail by the GUM approach. Uncertainty analysis was carried out on tensile testing to determine the traceability of the data obtained. Tensile testing produces data in the form of tensile strength and elongation of the test material. Tensile strength is a quantity that involves the magnitude of the force and length. Elongation is the increase in length when pulling. The tensile strength of the ring-shaped sample will be affected by the cross-sectional area on both sides of the ring when a tensile stress is applied. So, to determine the uncertainty of the tensile test, an analysis of the contributors to the uncertainty is required. This paper provides an explanation of how to analyze the uncertainty of the measurement of tensile strength and elongation in a ring-shaped sample. In addition, the value of the uncertainty in the measurement of tensile strength and elongation of SS304 (in the form of a ring) will also be presented. The test results show that the average tensile strength of SS304 rings is 631.070 ± 26.976 Nmm-2 with an expanded uncertainty of 50.451 Nmm-2 at a 95% confidence level. In addition, at the same level of confidence, the average SS304 ring elongation result is 18.681% ± 2.425% with an expanded uncertainty of 2.867%. Data repetition or repeatability is the biggest contributor to the uncertainty in measuring the tensile strength and elongation of the SS304 ring.
OPTIMASI PARAMETER METODE PEMISAHAN RADIONUKLIDA 95Zr DALAM LARUTAN PELAT ELEMEN BAKAR U3Si2/Al TERIRADIASI Yanlinastuti Yanlinastuti; Noviarty Noviarty; Iis Haryati; Ariyanti Saputri; Boybul Boybul; Supardjo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (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.2021.27.3.6535

Abstract

Radionuclide of 95Zr is one of the γ-radiation emitter radionuclides which has a short half-life so that it can be used as an indicator of burn up. In the irradiated U3Si2/Al fuel element plate there are several radionuclides of fission products emitting α, β and γ and when measuring radiation using γ-spectrometry, these radionuclides may interfere with each other resulting in inaccurate measurement results. Therefore, it is necessary to separate 95Zr for better accuracy in the measurement of γ-radiation with a γ-spectrometer. The purpose of this study was to obtain the optimum parameters for the separation of 95Zr in U3Si2/Al fuel element plate solution by ion exchange column method using Dowex 1x-8Cl- resin of 100-200 mesh in diameter. The optimum parameters obtained were then used for the separation of 95Zr in a solution of irradiated U3Si2/Al fuel element. It was found that the optimum parameter of acidity of H2SO4 in the feed was a concentration of 0.5 M, and the acidity of H2SO4 for elution was 1.0 M with a recovery of 98.20%. The weight of the resin to bind 95Zr radionuclide was 2.0 g with a recovery of 78.76%. The required feed volume was 50 µL with a recovery of 96.90% and the optimized flow rate was 0.1 mL/minute with a recovery of 96.72%. The optimum values were used for the separation of the U3Si2/Al fuel element plate solution irradiated with the code of Bottom (B), Middle (M) and Top (T). The measurement results show that the mean recovery of separation was 95Zr for code B-1 = 75.078%; B-2= 81.401%; M1=76,850%; M-2=83.806%; T-1=84.433%; and T-2 = 81.728% with the acceptability of the CV repeatability value smaller than the CV Horwitz value.
PEMISAHAN CESIUM DALAM BAHAN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 BURN UP 60% POTONGAN BOTTOM Ariyanti Saputri; Sutri Indaryati; Iis Haryati; Yanlinastuti Yanlinastuti; Hanifah Rifaatul Mahmudah; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (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.2021.27.3.6571

Abstract

The separation of cesium in U3Si2/Al fuel with a density of 4.8 gU/cm3 and 60% burn up by cation exchange using zeolite from Lampung has been carried out. Cesium isotope is used as a burn-up indicator for irradiated nuclear fuel. The separation aims to obtain an accurate cesium weight so that the calculation of burn up by the destructive method can be carried out correctly. The fuel plate of U3Si2/Al was cut in duplicates at the bottom with a weight of B-1=0.036 g and B-2= 0.037g, respectively. The fuel plate of U3Si2/Al with this weight was then dissolved using 6N HCl and 6N HNO3 in 10 mL. The solution of U3Si2/Al was then pipetted 1 mL and transferred from the hot cell to R.135 and diluted to 25 mL. The solution was pipetted 50 µL into a vial in duplicates and 1000 mg of Lampung zeolite was added. Separation process of 134Cs and 137Cs from 235U was done by cation exchange for 60 minutes. The results obtained were isotopes 134Cs and 137Cs in the solid phase while uranium (238U, 235U, 234U) and other isotopes in the liquid phase. The weight of the isotopes of 134Cs and 137Cs was then measured using a spectrometer/g. The weight of the isotopes of 137Cs and 134Cs in solution at Bottom position was B-1=0.00003283 g/0.036g and B-1= 0.000000147 g/0.036 g for 137Cs and 134Cs respectively while another duplicate was B-2=0.00003290 g/0.037g for 137Cs and B-2=0.000000222 g/0.037g for 134Cs. It can be concluded that in the separation of cesium in 50 µL solution of U3Si2/Al at Bottom position by cation exchange using 1000 mg of zeolite from Lampung, good results were obtained.

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