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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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STUDY ON CRITICALITY AND NEUTRONIC SAFETY PARAMETERS OF NUSCALE FUEL ASSEMBLY Zuhair; Wahid Luthfi; R. Andika Putra Dwijayanto; Suwoto
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.2021.27.3.6509

Abstract

NuScale, a typical Pressurized Water Reactor (PWR) Small Modular Reactors (SMRs), offers a new opportunity for the future of nuclear industry. With 160 MW thermal power, NuScale has several advantages such as flexibility due to its modularity in construction. This work is focused on the study of criticality and neutronic safety parameters of NuScale fuel assembly using MCNP6 code and ENDF/B-VII library. The calculation results shows that criticality of fuel assembly type D is the highest among other assembly types because it has a fuel pin with pure UO2 without Gd2O3 concentration. The Doppler temperature coefficient (DTC) of fuel assembly type C is the most negative among other assemblies due to Doppler broadening effect on resonance region of capture cross section of 238U which is the highest concentration. The moderator temperature coefficient (MTC) of fuel assembly type D is the most negative among the other assembly types. The effective delayed neutron fraction (βeff) does not reflect a consistent trend among fuel assembly types. Fuel assembly type D shows the highest prompt neutron lifetime (ℓ) while the highest neutron generation time (Ʌ) is shown in assembly type C. It can be concluded that this study provides adequate results that can be used as a first step to carry out the neutronic computation and analysis of the NuScale full core.
IDENTIFIKASI FASA PADA PELET BAHAN BAKAR U-ZrHx HASIL PROSES SINTER DENGAN ATMOSFER NITROGEN Masrukan; Jan Setiawan; Dwi Biyantoro
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): 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.2017.23.3.3641

Abstract

Phases identification of U-ZrHX fuel pellets sintered under nitrogen atmosphere has been carried out. This sintering process is intended to obtain higher density pellets that meet the requirements for use as a fuel in pressurized water reactor (PWR). The results of this study are expected to be used for the determination of phase and density changes occurred in U-ZrHx pellets at various Zr concentrations after sintering at various temperatures. PWR fuel from U-ZrHx pellet needs to be sintered to increase its density. At first, pellets made from U-ZrHx powder with Zr composition of 35, 45 and 55 wt.% through a pressing process at a pressure of 509-637 MPa. The resulted U-ZrHx green pellets were subsequently sintered under nitrogen gas atmosphere at 1200 °C and 1300 °C for a fixed sintering time of one hour. Phase analysis of the sintered U-ZrHx pellets was done by X-ray diffraction (XRD) technique and the data obtained were processed by using HighScore Plus software. The test results with XRD indicate the formation of phases such as U(ZrHr2), UO3, UO2, and Zr. The UO2 and Zr phases were characterized for their density, lattice parameter and percentage of weight. In this experiment it is desired to obtain U-ZrHx with high density without the formation of many phases. Density measurement shows that the density of zirconium metal and UO2 decreases with increasing sintering temperature (from 1200 °C to 1300 °C), while the lattice parameter increases with increasing sintering temperature. Data calculation of weight percentage for sintering at 1300 °C shows that the percentage values of the resulted zirconium metal and uranium dioxide are the closest to their theoretical calculation results.
ANALISIS CACAT BLISTER PADA KELONGSONG BAHAN BAKAR U3Si2/Al MENGGUNAKAN ULTRASONIC TEST Muhammad Fauzi; Roziq Himawan; Helmi Fauzi Rahmatullah; Setia Permana; Antonio Gogo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): 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.2017.23.3.3629

Abstract

Fuel integrity is one of safety considerations during nuclear reactor operation. Any defect resulted after reactor operation should be avoided. Thus, in nuclear fuel development, post-irradiation examination is necessary to be performed to ensure fuel integrity. Non-destructive test using ultrasonic method is one of alternative methods to perform post-irradiation examination. The objective of this analysis is to detect blister defect in cladding of plate type nuclear fuel. In post irradiation examination (PIE) using ultrasonic method, the operation parameters are necessary to determine. In this experiment, this analysis method was conducted on U3Si2/Al fuel element cladding consisting of blister defect. Ultrasonic tests were done using two techniques, i. e. pulse-echo and through transmission techniques. The results were presented in A-scan display. The results show that the longitudinal wave speed in water used in the testing is 1545 m/s. The optimal distance between the probe and the specimen when using pulse-echo technique is 20 mm. The optimal distance between the two probes when using through transmission technique is 40 mm. Tests at blister area using pulse-echo technique caused output signal to decrease by 19 %FSH, while through transmission technique caused output signal to decrease by 80 %FSH. According to the results it can be concluded that ultrasonic test is effective to detect a blister defect in plate type nuclear fuel, and through transmission technique exhibited a clearer difference between normal area and blister area in fuel cladding when compared to pulse-echo technique.
KARAKTERISTIK DAKTILITAS SS304 YANG TEROKSIDASI PADA TEMPERATUR TINGGI Jan Setiawan; Sungkono
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): 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.2017.23.3.3645

Abstract

This paper shows changes in hardness, tensile strength and ductility characteristics of SS304. The SS304 was processed by heat treatment at high temperature in atmospheric environment. SS304 is a type of stainless steel with many superior properties used in industry. In this research, a pipe made of SS304 with 30 mm in diameter was cut in transversal direction at 3 mm width to be used as a sample. The samples were processed by heat treatment at 600, 800 and 1000 °C for 1 and 2 hours in atmospheric environment. After the oxidation process, the samples were cooled by natural cooling outside the furnace or quenched in water. Hardness testing and ring tensile testing were done to the high-temperature oxidized SS304. Analyses were done to study changes in hardness, tensile strength and ductility characteristics of SS304 after oxidation process. The results show that hardness escalation occurs mainly for samples that were cooled by quenching in water compared to natural cooling. The highest hardness value was observed for the sample oxidized at 1000 °C for 1 hour and cooled by quenching, where hardness value changed from 281.25 HV before oxidation to 468.00 HV after oxidation. The tensile strength of the samples, however, tend to decrease, while the ductility of some samples tends to increase. Increase in tensile strength followed by increase in ductility occurs for samples oxidized at 600 °C for 2 hours with natural cooling. The tensile strength after oxidation reaches a value of 744.04 MPa with 12.38 % elongation escalation. As a guidance for implementation, SS304 in high temperature oxidation environment should be limited by temperature of 600 °C.
PENGARUH HIDROGEN TERHADAP KEKUATAN TARIK KELONGSONG BAHAN BAKAR NUKLIR BERBASIS PADUAN ZIRKONIUM Rohmad Sigit; Fajar Al Afghani; Jan Setiawan; Sungkono
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): 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.2017.23.3.3648

Abstract

During reactor operation, zircaloy-4 (Zr-4) as fuel cladding is under an environment that can potentially cause mechanical failure, especially by the effect of cladding-hydrogen interaction. Hydrogen would diffuse and precipitate to form brittle hydrides when the solubility limit is surpassed. In this experiment, ring tensile test method was used to examine the effect of hydrides existence on tensile strength of Zr-4 cladding. The Zr-4 sample was annealed at 500 °C and 600 °C for an hour in vacuum condition and followed by hydriding with a pressure up to 1000 mbar with 3 hours holding time and slow cooling. According to ideal gas equation, hydrogen content in Zr-4 reaches about 674 ppm and 1215 ppm at hydriding temperature of 500 °C and 600 °C respectively. Microstructure observation shows that hydrides precipitate was formed at grain boundaries. The hydrides were found to be longer and thicker at higher hydriding temperature. As the hydrogen content increases, the ring tensile test shows increasing stress at 4.4 %, decreasing strain at 2.1 % and increasing microhardness level at 13 %. This result confirms that the existence of hydrides precipitates in cladding has an effect of increasing the tensile strength and decreasing the ductility obviously.
ANALISIS KOROSI PADUAN ZIRLO-Mo DALAM MEDIA NaCl MENGGUNAKAN METODE POLARISASI Maman Kartaman Ajiriyanto; Dian Anggraini; Rosika Kriswarini
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): 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.2017.23.3.3221

Abstract

Zirlo-Mo is a zirconium type of alloy, which is developed as cladding material for light water reactor. As cladding material, the important parameter to be characterized is its corrosion resistance properties. The goal of this research is to study the corrosion resistance of zirlo-Mo in NaCl medium with varied concentration of 0.03; 0.3; and 3 % and compare it to the those of zircaloy-2 dan zircaloy-4 in the same medium. Samples were mounted into disk form and metallographically prepared, i.e. treatment with abrasive paper of 500 to 1200 grit SiC. The analysis of zirlo-Mo corrosion rate was performed by polarization method with a potentiostat. The advantage of this polarization resistance method is that it takes shorter testing time if compared to weight gain method, and it can be applied to analysis at low corrosion rate of below 0.1 mpy. Other tests were also done to support the corrosion rate analysis such as microhardness test with microhardness tester and microstructure examination with optical microscope. The tests results show that in general zirlo-Mo, zircaloy-2 and zircaloy-4 possess very low corrossion rate, i.e. below 1 mpy. The corrosion rate of zircaloy-2, zircaloy-4 and zirlo-Mo in NaCl 3 % medium are 0.006; 0.007 and 0.590 mpy respectively. The corrosion rate of zirlo-Mo is relatively the highest due to its lath martensite or accicular structure, which has high hardness value. The lath martensite structure and high hardness properties make zirlo-Mo highly susceptible to corrosion. Hardness test results show that the hardness value of zircaloy-2, zircaloy-4 and zirlo-Mo are 194.79; 227.95 and 360.05 HV respectively. The existence of solid solution of Mo and Zr-α as well as the martensite or acicular structure of zirlo-Mo result in high hardness value of the alloy.
PENGARUH PENAMBAHAN ION SULFAT TERHADAP REKOVERI PEMISAHAN URANIUM MENGGUNAKAN METODE KOLOM PENUKAR ANION Boybul Boybul; Yanlinastuti Yanlinastuti; Dian Anggraini; Aslina Br. Ginting; Arif Nugroho; Rosika Kriswarini
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 3 (2017): 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.2017.23.3.3679

Abstract

In the previous research, it was observed that the recovery of uranium separation by anion exchange column method with HCl 12 N solvent was 68.2137 %. The recovery can be increased by optimizing the uranium separation parameters by the use of HCl and Na2SO4 solvent mixture. A 500 μL uranyl nitrate standard solution was used as feed with varied HCl concentration of 6 N, 9 N, 12 N and varied Na2SO4 concentration of 500, 1000, 1500, 2000, 2500 and 3500 ppm as reagent solution. In the anion exchange process, uranium was bound by the resin to form [UO2(SO4)3]4- and [UO2(SO4)2]2 complex compounds, and these complexes were eluted with 15 mL of 0.15 N HCl dilute solution. The eluate was heated to nearly dried condition and 1 mL of HCl concentrated solution was added for electrodeposition process. The uranium deposited on stainless steel planset, as the result of electrodeposition process, was analyzed with alpha spectrometer. It was observed that the largest 235U content was found in 9M of HCl solvent and 1500 ppm of Na2SO4 with 95.7124 % recovery. Therefore, the concentration value of HCl of 9 N and 1500 ppm of Na2SO4 were subsequently applied to the uranium separation of irradiated U3Si2/Al fuel element plate. An amount of 500 μL supernatant of U3Si2/Al specimen taken from the top, middle and bottom position of fuel element plate (FEP) were put into an anion exchange column containing 1 g of Dowex-Cl resin. The results of the separation process were 235U isotope identified according to their position in the irradiated U3Si2/Al fuel element plate i.e. top, middle, and bottom. It was found that the recovery values were 13265.7866 μg/gFEP; 6923.9248 μg/gFEP; and 12332.4175 μg/gFEP representing their position of top, middle and bottom respectively. It is concluded that the use of 9M of HCl solvent and 1500 ppm of Na2SO4 by anion exchange column method can improve uranium separation recovery from 68.2137 % to 95.7124 %.
PENGARUH VISKOSITAS SOL DAN PRESOLIDIFIKASI PADA PROSES GELASI EKSTERNAL DALAM PEMBUATAN GEL AMMONIUM DIURANAT MENGGUNAKAN SURROGATE CERIUM Sukarsono; Meniek Rachmawati
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): 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.2017.23.2.3229

Abstract

The external gelation process is used in the preparation of ammonium diuranat gel with processed into UO2 coated particles. UO2 coated particles are the core of high temperature reactor (HTR) fuels. This paper reported the research of gel preparation though external gelation process using a surrogate of zirconium stabilized cerium. The purpose of this study was to determine the effect of sol viscosity and presolidification related to gel roundness which was produced in the operation of the gelation unit in the laboratory of High Temperature Nuclear Reactor Fuel. The parameters studied were sol viscosity by varying the addition of additive and gelation process with presolidification (with NH3 gas stream) and without presolidification (without NH3 gas flow). The Ce-Zr gel can be prepared through a similar process of ADU gel preparation. A mixture of cerium and zirconium nitrate solution with Ce content of 12 mole %, was added by urea as a pH regulator and added by additive materials of tetra hydrofurfuryl alcohol (THFA) and poly vinyl alcohol (PVA). The solution was ready to be used in the gelation process. Sol solution was fed into the gelation column by pressing using compressed air. The sol solution entered the gelation column through the dripping node nozzle was vibrated with frequency of 90-110 Hz and certain amplitude. The droplets was passed on a column containing NH3 gas so that occurred presolidification on the surface of the droplets. The drops fall to the solution medium containing NH4OH in the gelation column. The compaction occurred due to the polymerization of cerium nitrate and zirconium nitrate at a pH of about 10. The gel was accommodated in a container located under the gelation column and allowed to be at least 2 hours to complete the solidification. The gel was washed successively with aqueous NH4OH, water and isopropyl alcohol and continued by drying. Operation of the gelation column to produce a round and uniform gel required adjustment the frequency, amplitude and flow rate of the sol solution. Operation of gelation columns at a frequency of 100 Hz using a sol with a viscosity of 97 cP and a flow rate of 25.8 mL/min produced a gel with smooth, uniformity and a relatively good sphere. For sol with a viscosity of 56 cP had relatively good flow rate was 28.8 mL/min and at a low viscosity of 46 cP, a flow rate of 30 ml/min is obtained. Operating the gelation device at frequencies of 110 Hz, 100 Hz and 90 Hz, produced a smooth, uniform and round gel at a sol flow rate of 27.9 mL/men, 25.8 mL/min and 21.2 mL/min. The gelation process with NH3 gas stream as a presolidative medium produced a more rounded and uniform gel than with no NH3 gas stream that resulted the non-round gel. From this studied it could be concluded that the viscosity of the sol and the vibration frequency of the dropout nozzle affects the flow rate of the sol but did not affect the gel roundness. The larger viscosity of solution, the smaller the flow rate of the sol to get a separate, uniform and round gel. The greater vibrational frequency also increases the flow rate to get a good gel. The gelation process without presolidification resulted a non-round gel.
KARAKTERISTIK INGOT PADUAN U-Zr-Nb PASCA PROSES QUENCHING Masrukan; Jan Setiawan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): 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.2017.23.2.3522

Abstract

The U-Zr-Nb heat-quenching experiment has been subjected to various Nb compositions. Ingots of U-Zr-Nb alloys with varying Nb compositions are 1%, 4% and 7% and a fixed 10% Zr composition is made by melting U, Zr and Nb metals in an electric arc furnace. The quenching experiment is intended to convert an unstable phase αU into stable γU phase. Quenching experiment is intended to convert an unstable phase αU phase into a stable γU phase. In quenching experiments, the U-10Zr-1Nb, U-10Zr-4Nb, U-10Zr-7Nb ingot ingredients were fed into box that made of steel plate and then fed into a heating furnace and heated at 900 C and held for 2 hours for then cooled rapidly in water. After quenched then subjected to phase testing and microstructure observation. The phase test is performed using X-ray diffraction (XRD) and the data is processed using HighScore software, while the microstructure observation using optical microscope. Test results with XRD showed that the U-10Zr-1Nb alloy before quenched consisted only of an αU phase, the U-10Zr-4Nb alloy consist of 23,1504% αU phases and 76,8495% γU phases, while U-10Zr-7Nb consist of 34,1873% αU phases and 65,8127% γU phases. For U-10Zr-1Nb alloys after quenched consist of an αU phase, the U-10Zr-4Nb alloy consist of 44% αU phases and 46% γU phases, the U-10Zr-7Nb alloy consisted of 19% αU and 81% phases γU. For U-10Zr-1Nb alloys after quenched consist of an αU phase, the U-10Zr-4Nb alloy consisted of 44,6711% αU phases and 55,3289% γU phase, the U-10Zr-7Nb alloy consist of 17,9918% αU phases and of 82,0082% γU phases. The results of the theoretical density analysis on the phases formed on U-10Zr-1Nb, U-10Zr-4Nb, U-10Zr-7Nb alloys both before and after quenched were obtained that ingot of U-10Zr-1Nb, U-10Zr-4Nb alloy U-10Zr-7Nb shows that the αU and γU phase densities decrease slightly if the Nb content is lower. Meanwhile, microstructure observation results showed that in the higher of Zr content (7% Nb) will form smaller grains (fine) after quenching. It can be concluded that the quenching process of U-10Zr-1Nb, U-10Zr-4Nb and U-10Zr-7Nb alloys will change the phase that is from the αU phase to γU and the change the microstructure formed. The microstructural changes show that in the higher Nb content, the grains that are formed become finer.
PENGARUH PELARUT ORGANIK PADA PROSES PERTUKARAN ANION DALAM PEMISAHAN URANIUM DARI LARUTAN PEB U3Si2/Al PASCA IRADIASI Dian Anggraini; Boybul Boybul; Yanlinastuti Yanlinastuti; Arif Nugroho; Rosika Kriswarini; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): 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.2017.23.2.3576

Abstract

Separation method of uranium from nuclear fuel solutions was developed to improve the quality of uranium-based analysis of post-irradiation nuclear fuel. The aim of this study was to determine the effect of organic solvent on the anion exchange process on the recovery of U from U3Si2/Al fuel solution post-irradiated. The organic solvent, in this case methanol, contained in solvent medium in the anion exchange column may be possible to increase the resin sorption of the uranium. The materials used are standard solution of uranil nitrat and supernatant solution of Cs separation process from the fuel solution. The uranium separation method from U3Si2/Al post irradiated fuel solution uses two columns of anion exchanger with two stages. The first column uses Dowex 1x8-Cl Dowex resin. Parameters varied in the first column are HNO3 concentrations ranging from 1M; 2M; 3M and 4M. The parameters variation in the second column is the ratio of HCl/methanol volume of 30:70 %; 20:80 % and 10:90 %. The result obtained a maximum U recovery of 90,68 % with feed solution composition in the first column containing 3 % HNO3 and methanol by 50 %volume, while in the second column with 6M HCl composition with methanol in 10 %:90 % ratio. In these conditions obtained uranium content in a solution of the middle PEB U3Si2/Al TMU 2.96 g/cm3 piece of 0.4725 μg.

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