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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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FENOMENA KOROSI ZIRKALOY-2 DAN ZIRKALOY-4 DALAM MEDIA LARUTAN NaCl SECARA ELEKTROKIMIA Dian Anggraini; Maman Kartaman Ajiriyanto; Rosika Kriswarini; Yanlinastuti Yanlinastuti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 21 No. 3 (2015): 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.2015.21.3.2465

Abstract

Zircaloy-2 and zircaloy-4 are used as cladding material in Light Water Reactor, (LWR), Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR). During its use in the spent fuel pool may interact between the cladding and seawater in case of abnormal condition. Chloride ion which contained in seawater has the potential for being corrosion in cladding material. The aim of this work was knowing zircaloy-2 and zircaloy-4 corrosion phenomena in NaCl medium by corrosion parameter determination with polarization resistance technique, tafel, and potentiodynamic. Samples are zircaloy-2 and zircaloy-4 pieces material with 1x1 cm dimension. The samples are mounting and soldered with copper wire, then the sample surface are polished by 1200 grade grinding paper. Corrosion test was done with corrosion cell that was completed by electrode standard (saturated calomel), electrode buffer (graphite) and work electrode (sample). Solvent medium was NaCl solution with concentration 3.5 %; 0.35 %; and 0.175 %. The result showed that ion Clˉ contained in NaCl which were concentration area from 0.175 % to 3.5 % did not affect corrosion mechanism. The corrosion type in potential range ± 250 mV to E­­corr was uniform corrosion in ZrO­2 oxide form. Zirkaloy-2 and zirkaloy-4 corrosion rate in 3.5 % NaCl concentration were obtained 6.39×10-3 and 7.40×10-3 mpy respectively. Corrosion phenomenon which was observed by potentiodynamic technique (± 1000 mV to Ecorr) showed that zirkaloy-2 and zirkaloy-4 underwent passivation and pitting corrosion. Pitting potential zirkaloy-2 and zirkaloy-4 were obtained -452.8 mV and -182.8 mV severally.
FABRIKASI MIKROSFIR UO2 MENGGUNAKAN TEKNIK AERASI Meniek Rachmawati; Etty Mutiara; Tri Yulianto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): 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.2016.22.2.3089

Abstract

It has been developed a fabrication process of low density UO2 microspheres for direct feed in pelletization process of PHWR advanced reactor fuel. The fabrication has been implemented by a sol-gel method using aeration technique on the sol / broth with external methods and three variations of gelation. In the aeration technique, broth is directly prepared in gelation mode without one night settling time. The aeration technique is the opposite of the deaeration technique, which is used in fabricating microspheres HTGR fuel. The broth which has been prepared with a mole ratio of NO3 / U between 1.5 to 1.7 with pH solution of 1.6 and viscosity between 630-660 cP directly to be gelated in three ways of gelation. The process of both gelation method 1 and method 2 is implemented by passing through the broth to a dispersion nozzle with 1 mm of diameter which is vibrated by an electromagnetic vibrator at 150 Hz with medium for free fall droplet differently before entering into a solution of NH4OH, while the gelation method 3 implemented manually. The wet UO2 microspheres derived from the three ways of gelation above are treated by heating process at the same way that is dried at a temperature of 85 ºC and 220 ºC each respectively for 1 hour, followed by a calcination process of microspheres UO2 for 1 hour at a temperature of 500 ºC in gaseous medium of O2 and a reduction process at a temperature of 600 ºC in gaseous mixture of N2 and H2 medium for 1 hour. The UO2 microspheres gelation of method 3 are chosen to be sorted and characterized. The characterization results provide the characteristics data of UO2 microspheres in the form of microspheres diameter of 900 μm, tap density of 1.90 g/cm3 and specific surface area of 6 m2/g. The results of the analysis and characterization are then compared to other data of other research so that it can be concluded that the use of such an aeration technique on the broth can produce low density UO2 microspheres that qualify as direct feed for fuel pelletization process of the advanced PHWR and advanced PWR. The fabrication using the aeration technique with gelation method 2 has the greatest opportunity to produce UO2 microspheres with a high production rate and certain characteristics if it is implemented by resetting the rate of dispersion and the free fall droplet duration.
KORELASI KOMPOSISI UNSUR TERHADAP SIFAT TERMAL SERBUK BAHAN BAKAR U-ZrHX Masrukan; M. Husna Alhasa; Yanlinastuti Yanlinastuti
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): 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.2016.22.2.3090

Abstract

Analysis has been conducted to determine the composition correlation on the thermal properties of the powder fuel U-ZrHx. U-ZrHx powder made from the process hydriding U-Zr ingot, where the ingot is the result of U-Zr and Zr U metal melting. In this experiment made three variations of powders, namely U-35ZrHx, U-45ZrHx, and U-55ZrHx. Need for determination of the thermal properties of Zr was to determine the effect of the nature of the transformation of Zr levels of heat from the fuel. At first, U and Zr metal is melted in electric arc furnaces to produce ingot U-Zr. U-Zr ingot then made powder with hydridring-milling techniques to produce U-Zr powder. U-Zr powder composition analyzed using techniques spectroscopy atomic absorption (AAS) and UV-Vis spectroscopy. The results of composition analysis showed that the analysis to determine the content of U and Zr nearly all the test samples analyzed have quite a big difference between the content of U and Zr as determined by the results of the analysis of U and Zr exception analysis result in powder U-45Zr which differ only 0.609%. From the analysis of impurities obtained that nearly all the impurities that exist still meet the requirements for fuel unless the elements Fe, where elements of the existing Fe amounted to 382.912 g/g while the requirement of £ 250 mg /µg. Testing conducted heat capacity in the temperature range 35 ° C to 437 ° C showed that the capacity were greatest powder 35ZrHx U-with a value heat capacity of 0.13 J / g.oC. Meanwhile, test results obtained transition phase change that the U-45ZrHx undergo two stages of reaction with phase change. It can be concluded when seen from the content of U and Zr cannot be used for fuel, while the analysis of the content of impurities found that all the elements that are still eligible for the fuel unless the element Fe. For the analysis of the thermal properties are the heat capacity of the heat capacity of the highest values obtained in powder U-35ZrHx, while the transition from the testing phase changes shows that the U-45ZrHx undergo two stages of reaction with phase change. There is an effect of the composition on the thermal properties, where the higher the content of Zr, the value of uranium zirconium hydride heat capacity is lower.
PHASE CHANGES ON 4H AND 6H SIC AT HIGH TEMPERATURE OXIDATION Jan Setiawan; Ganisa Kurniati Suryaman; Masrukan
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): 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.2016.22.2.3091

Abstract

The oxidation on two silicon carbide contains 6H phase and contains 6H and 4H phases has been done. Silicon carbide is ceramic non-oxide with excellent properties that potentially used in industry. Silicon carbide is used in nuclear industry as structure material that developed as light water reactor (LWR) fuel cladding and as a coating layer in the high temperature gas-cooled reactor (HTGR) fuel. In this study silicon carbide oxidation simulation take place in case the accident in primary cooling pipe is ruptured. Sample silicon carbide made of powder that pressed into pellet with diameter 12.7 mm and thickness 1.0 mm, then oxidized at temperature 1000 oC, 1200 oC dan 1400 oC for 1 hour. The samples were weighted before and after oxidized. X-ray diffraction con-ducted to the samples using Panalytical Empyrean diffractometer with Cu as X-ray source. Diffraction pattern analysis has been done using General Structure Analysis System (GSAS) software. This software was resulting the lattice parameter changes and content of SiC phases. The result showed all of the oxidation samples undergoes weight gain. The 6S samples showed the highest weight change at oxidation temperature 1200 oC, for the 46S samples showed increasing tendency with the oxidation temperature. X-ray diffraction pattern analysis showed the 6S samples contain dominant phase 6H-SiC that matched to ICSD 98-001-5325 card. Diffraction pattern on 6S showed lattice parameter, composition and crystallite size changes. Lattice parameters changes had smaller tendency from the model and before oxidation. However, the lowest silicon carbide composition or the highest converted into other phases up to 66.85 %, occurred at oxidation temperature 1200 oC. The 46S samples contains two polytypes silicon car-bide. The 6H-SiC phases matched by ICSD 98-016-4972 card and 4H-SiC phase matched by ICSD 98-016-4971 card. Diffraction pattern on 46S also showed lattice parameter, composition and crystallite size changes. The lattice parameter changes not significant. For 6S and 46S samples at 1400 oC, the 6H-SiC phase changes into other phases more than 50 % from its original weight percentage.
METODE PENGENDAPAN DAN PENUKAR KATION UNTUK PEMISAHAN CESIUM DALAM BAHAN BAKAR U3Si2-Al Aslina Br. Ginting; Dian Anggraini
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): 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.2016.22.2.3092

Abstract

137Cs isotope was one of the fission products which could be used as a burn up monitor to study the performance of the fuel during irradiation in a reactor. For analyzed 137Cs isotope in post-irradiation U3Si2-Al fuel element plate (PEB), it was needed a valid method to obtain accurate data. Several methods could be used for 137Cs separation from U3Si2-Al PEB solution were precipitation method in CsClO4 salt form according to ASTM E 320-79 and cation exchange method using Lampung zeolite. The precipitation process was done using CsNO3 powder as a carrier substance and HClO4 reagent, while the cation exchange process was done by the addition of Lampung zeolite. The objective research was getting a valid method for 137Cs separation in post-irradiation U3Si2-Al PEB, especially for the aspects of CsNO3 powder weight influence and weight of zeolite Lampung was added. 137Cs isotope precipitation process was done by pipetting 150 mL U3Si2-Al PEB solution which was added to the 500; 625; 700; and 1000 mg weight variation CsNO3 powder and 4 mL HClO4 in ice media for 1 hour. The precipitation process results were obtained CsClO4 precipitate separated from the supernatant as a liquid phase. Meanwhile, the cation exchange process was done by adding 700; 900; 1000 and 1200 mg weight variation Lampung zeolite by shaking for 1 hour. The cation exchange process results was obtained 137Cs-zeolite in the solid phase and the other isotopes in the liquid phase. The137Cs content in 137CsClO4 precipitate,137Cs-zeolite solid phase and supernatant was measured using spectrometer-g. The analysis showed that the optimum weight of CsNO3 addition when most binding of 137Cs occurred was 700 mg equal to 0.0472 g 137Cs isotope, while the optimum weight addition of Lampung zeolite was 1000 mg until 1200 mg equal to 137Cs isotope content in137Cs-zeolite obtained was 0.0557 g. The 137Cs isotope separation process using both methods could be concluded that the cation exchange method using zeolite by the addition of 1000 mg Lampung zeolite until 1200 mg had better result than the precipitation method. The other advantages of this method were easier and safer than precipitation methods that the process should be in ice batch (-4oC), using chemicals HClO4, acetone and ethanol which had volatile and exothermic character.
PENGARUH PARAMETER PROSES ELEKTRODEPOSISI TERHADAP PENENTUAN BERAT ISOTOP 235U DALAM PEB U3Si2/Al PASCA IRADIASI Yanlinastuti Yanlinastuti; Boybul Boybul; Aslina Br. Ginting; Dian Anggraini
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): 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.2016.22.2.3093

Abstract

Electrodeposition process for determination of 235U isotope is affected by current, contact time, feed volume and type of buffer solution. The aim of this research is to obtain optimum parameter of electrodeposition process for 235U determination in U3Si2/Al irradiated fuel plate. Electrodeposition process is done by some variation of parameters, that are the current 1; 1.2; and 1.4 ampere with variation of time 1; 1.5; 2.0; 2.5 and 3.0 hours. The standard solution volume of 20% U3O8 are 100; 300; 500; 700 µL with variation of electrolyte buffer solution NH4Cl, NH4OAc, NH4HSO4 dan (NH4)2SO4. The result of electrodeposition process showed that optimum condition electrodeposition process for 235U isotope was obtained currents 1.2 ampere with process time for 2 hours, and 300 µL volume feed with electrolyte buffer NH4HSO4 in pH 3,5. The results of the electrodeposition process was obtained 2.314 µg deposition of the isotope 235U with recovery of 99.91%. The condition optimum parameter above was used for electrodeposition process in determination of 235U isotope which contained in U3Si2/Al irradiated fuel element. The spectrometer alpha analysis was obtained weight of 235U isotope as 0.403 µg with recovery of 91.80%. The conclusion of this research, the electrodeposition methods is good enough to be used for determining 235U isotope in U3Si2/Al irradiated fuel plate.
PEMBUATAN RADIONUKLIDA MOLIBDENUM-99 (99Mo) HASIL AKTIVASI NEUTRON DARI MOLIBDENUM ALAM UNTUK MEMPEROLEH TEKNESIUM-99m (99mTc) Indra Saptiama; Herlina; Sriyono; Endang Sarmini; Abidin; Kadarisman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 2 (2016): 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.2016.22.2.3094

Abstract

Uranium usage restriction causes the hospitals in Indonesia difficult to obtain the supply of 99mTc. At Present, 99mTc is obtained from molybdenum as a uranium fission product. Production of 99Mo radionuclide resulted from neutron activated natural molybdenum (MoO3) in G.A Siwabessy reactor could be used as an alternative method for producing 99mTc. The aim of this research is synthesized of 99Mo radionuclide from neutron activated natural molybdenum (MoO3) to obtain 99mTc. The five grams of MoO3 powder was packed in a quartz ampule and inserted into inner capsule then also inserted into outer capsule as a target material. It was irradiated in G.A Siwabessy reactor for 100hours. Based on theoritical calculation, about 65 % of maximum 99Mo activity could be recovered. After irradiation, MoO3 was dissolved by NaOH 4 M solution, so it was natrium molybdate (Na2MoO4) solution. 99Mo and 99mTc radionuclide were analyzed using gamma spectrometer. 99Mo radionuclide was detected on Na2MoO4 solution as product that had specific activity of 0.81 Ci 99Mo/ g Mo. 99mTc as daughter radionuclide was separated from 99Mo as parent radionuclide using separated column containing zirconium-based material (ZBM) as 99Mo adsorbent. 99mTc radionuclide has been successfully separated using ZBM column although recovery of 99mTc was quite low in which approximately 52 to 71 %. The 99mTc radionuclide was recovered in the form of sodium pertechnetate (NaTcO4) solution.
PENGARUH TEMPERATUR DAN IRADIASI TERHADAP INTERDIFUSI PARTIKEL BAHAN BAKAR JENIS U−7Mo/Al Maman Kartaman Ajiriyanto; Aslina Br. Ginting; Supardjo; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): 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.2016.22.1.2740

Abstract

U−7Mo/Al alloy had great potential as research reactor fuel, but it had several disadvantages, such as, it can form a interaction layer at the interface during the process of fabrication and irradiation in a reactor. The research objective was determining the interaction layer that was caused by interdiffusion or diffusion couple of U−7Mo with AlMg2 alloy which was annealed at 500 °C and 550 °C for 24 hours. The observation of microstructure used a Scanning Electron Microscope (SEM) on diffusion couple sample which was heated at temperature of 500 °C had not seen the layer interaction or the formation of a new phase between particles of U−Mo and Al matrix, but heating treatment at 550 °C indicated that U−7Mo alloy had been interdiffusion with AlMg2 plate produced (U,Mo) Alx compound on the interfaces. It was evidenced by interdiffusion reaction analysis used DTA that showed that U−7Mo / Al alloy at 500 °C had good heat compatibility, but at temperatures upper than 550 °C it had been phase changed from a + d to a + g phase. The heating in DTA furnace up to 679.14 °C produced U(Al,Mo)x meta stable phase and then interdiffusion process with uranium molten formed layer interaction that formed UAlx compound agglomerates (UAl4, UAl3 and UAl2). Agglomerates was formed from the heating process which was similar to agglomerates that caused by irradiation. U−7Mo / Al Fuel alloy that had 58% burn up had been interdiffusion between U−7Mo with Al matrix produced U(Al,Mo)x metastable phase that turned into (U, Mo) Al7 layer, UMo2Al20 precipitates, (UMo)Al3−Al and formed a boundary or UAlx (UAl4, UAl3 and UAl2) agglomerates. The results of microstructure analysis used SEM and interdiffusion reactions used DTA was supported by the analysis of micro hardness used Vickers Hardness. The results of hardness analysis that was done to AlMg cladding and U−7Mo alloy (before and after heating) and diffusion couple of U−7Mo / Al samples with AlMg2 plate after heating at 550 °C were respectively 64.62 and 340.45 HV (before heating) and 52.34; 303.16 and 497.34 HV (after heating). Diffusion couple U−7Mo/Al with AlMg2 plate samples had the highest hardness value. This hardness difference showed that the interdiffusion test used diffusion couple produced a new compound (U, Mo) Alx in interface zone that had different character, but the formation of interaction layer is not expected in the fuel U−Mo / Al dispersion because micro hardness and density of (U, Mo) Alx compound’s layer was lower than the average density of U−7Mo/ Al alloy.
PENGARUH UNSUR Zr TERHADAP PERUBAHAN SIFAT TERMAL BAHAN BAKAR DISPERSI U-7Mo-xZr/Al Supardjo; Agoeng Kadarjono; Boybul Boybul; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): 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.2016.22.1.2741

Abstract

Thermal properties data of nuclear fuel is required as input data to predict material properties change phenomenon during the fabrication process and irradiated in a nuclear reactor. Study the influence of Zr element in the U-7Mo-xZr/Al (x = 1%, 2% and 3%) fuel dispersion to changes in the thermal properties at various temperatures have been stiffened. Thermal analysis includes determining the melting temperature, enthalpy, and phase changes made using Differential Thermal Analysis (DTA) in the temperature range between 30 °C up to 1400 °C, while the heat capacity of U-7Mo-xZr alloy and U-7Mo-xZr/Al dispersion fuel using Differential Scanning Calorimeter (DSC) at room temperature up to 450 °C. Thermal analyst data DTA shows that Zr levels of all three compositions showed a similar phenomenon. At temperatures between 565.60 °C – 584.98 °C change becomes α + δ to α + γ phase and at 649.22 °C – 650.13 °C happen smelting Al matrix Occur followed by a reaction between Al matrix with U-7Mo-xZr on 670.38 °C – 673.38 °C form U (Al, Mo)x Zr. Furthermore a phase change α + β becomes β + γ Occurs at temperatures 762.08 °C – 776.33 °C and diffusion between the matrix by U-7Mo-xZr/Al on 853.55 °C – 875.20 °C. Every phenomenon that Occurs, entalphy posed a relative stable. Consolidation of uranium Occur in 1052.42 °C – 1104.99 °C and decomposition reaction of U (Al, Mo)x and U (Al, Zr)x becomes (UAl4, UAl3, UAl2), U-Mo, and UZr on 1328,34 °C - 1332,06 °C , The existence of Zr in U-Mo alloy increases the heat capacity of the U-7Mo-xZr/Al, dispersion fuel and the higher heat capacity of Zr levels increased due to interactions between the atoms of Zr with Al matrix so that the heat absorbed by the fuel increase.
PENGARUH PROSES SINTERING TERHADAP PERUBAHAN DENSITAS, KEKERASAN DAN MIKROSTRUKTUR PELET U-ZrHx Masrukan; Mujinem
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): 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.2016.22.1.2742

Abstract

The process of sintering fuel pellets U-ZrHx performed to obtain a higher density compared pellets before sintered. At first, U-ZrHx pellets made from the content of Zr respectively 35%, 45% and 55% by weight. Sintering process is carried out for 1 hour at varying temperature 1100°C, 1200°C and 1300°C with an increase rate of 2°C temperature/ min, then lowered the temperature at a rate of decrease in temperature of 10°C / min until room temperature. U-ZrHx pellets that have been sintered subsequently subjected to various tests, such as: the dimensions, density, hardness, and microstructure. In testing the density, it appears that at the same Zr composition, but the sintering temperature is higher than an increase in the density. When seen in the same sintering temperature and Zr content is higher then the lower the density obtained. Micro hardness testing results show that the composition Zr 35% fixed and 55% were sintered at the sintering temperature is higher, have the higher the hardness, and then decreases as the sintering temperature increase. If the test results viewed in the sintering temperature fixed but the content of Zr is different, it can be seen from the content of Zr 35% (U-35ZrHx) to 45% (U-45ZrHx) an increase in hardness and will decrease to an increase in the content of Zr to 55% (U- 55ZrHx). Meanwhile, the results of the microstructure observation of the pellets can be seen that at composition Zr fixed but the sintering temperature is higher, then the distance between the powder / porosity increasingly dense. When viewed from the sintering temperature fixed but the composition increasingly higher Zr it is seen that higher Zr composition, the density between the powder in the pellets decreased. It can be concluded that the sintering process the pellets will increase density, hardness, and distance between powder increasingly dense. The optimum condition is achieved at sintering process the pellets with a composition of 45% Zr (U-45ZrHx) sintered at temperatures of 1200°C. In these conditions the sintered the pellets has a density of 8.673 g/cm3, hardness of 661 HVN without developing cracks.

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