Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
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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<![CDATA[ANALISIS KANDUNGAN CESIUM DAN URANIUM DALAM BAHAN BAKAR U3Si2/Al PASCA IRADIASI ]]>
<![CDATA[Boybul Boybul]]>;
<![CDATA[Yanlinastuti Yanlinastuti]]>;
<![CDATA[Arif Nugroho]]>;
<![CDATA[Dian Anggraini]]>;
<![CDATA[Rosika Kriswarini]]>;
<![CDATA[Aslina Br. Ginting]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.2.3568
<![CDATA[Separation of fission product isotopes of 137Cs and 235U has been done to determine content of 137Cs and 235U in irradiated U3Si2/Al fuel plate of 2.96 gU/cm3 density. The initial step in the determination of 137Cs and 235U content is sample cutting process. The cutting position was determined at the top middle and bottom part of the plate by an amount of 0.095 g; 0.1103 g and 0.1441g for each respective position. Each sample was dissolved in 25 mL of HNO3 6N dan HCl 6N to obtain irradiated U3Si2/Al solution. Separation was carried out by cation exchange and precipitation methods. The separation of 137Cs was done by pipetting 150 µL of the irradiated U3Si2/Al sample solution and adding it to 1000 mg of Lampung zeolite, while the separation by precipitation method was done by adding 700 mg of CsNO3 powder as carrier and HClO4 to the U3Si2/Al sample solution. In the cation exchange method, 137Cs would be bound by zeolite in the solid phase in the form of 137Cs-zeolite and uranium would stay in the liquid phase or supernatant. The precipitation was done in an ice bath at -4 oC to obtain 137CsClO4 precipitate. The effluent leaving the column was subjected to electrodeposition process for 235U content measurement with α-spectrometer. The 137Cs content obtained by cation exchange was 401.0335 µg/g FP (FP = irradiated fuel plate); 451.1094 µg/g FP and 343.9651 µg/g FP for each respective sample position by 99 % recovery, while the 137Cs content obtained by precipitation method was 393.4581 µg/g FP; 452.0525 µg/g FP and 330.7839 µg/g FP for each respective sample position by 98 % recovery. The 235U content was found to be 45208 µg/g FP; 50896 µg/g FP and 44336 µg/g FP for each respective sample position by 68 % recovery.]]>
<![CDATA[PENGARUH DENSITAS URANIUM TERHADAP UMUR BAHAN BAKAR NUKLIR DI DALAM REAKTOR RSG-GAS DITINJAU DARI ASPEK NEUTRONIK ]]>
<![CDATA[Saga Octadamailah]]>;
<![CDATA[Supardjo]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.2.3550
<![CDATA[Material Testing Reactor (MTR) is a research reactor with characteristics: small power, high thermal neutron flux and short life time fuel. Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) located in Serpong is an MTR type reactor initially operated with dispersion fuel of U3O8/Al with uranium enrichment of 19.75 % 235U and uranium density of 2.96 gU/cm3. In an effort to increase the lifetime of nuclear fuel, BATAN has done research on high density uranium fuel starting from U3Si2/Al fule and continued with UMo based fuel in line with investigations performed by worldwide researchers in the field of research reactors. Fuel of U3Si2/Al with a density of 2.96 gU/cm3 has been successfully manufactured and utilized as fuel for RSG-GAS to substitute the U3O8/Al fuel. Post irradiation examination has also been done for U3Si2/Al with a density of 4.8 gU/cm3 up to a burn up of 60 %, while research on UMo with 7 gU/cm3 density has been progressing into mini plate fabrication. Research on high density fuel has been focused on fabrication process, while little has been done on lifetime calculation or utilization period. Therefore, this research has performed a calculation on nuclear fuel lifetime using a paired program of ORIGEN and MCNP. The calculation results show that an increase in uranium density will extend the lifetime of fuel exponentially. The longest lifetime is achieved by UMo type, which extends to 651 days, and this relates to a lifetime extension of 219 days if the RSG-GAS is operated with UMo fuel.]]>
<![CDATA[ANALISIS KRITIKALITAS BAHAN BAKAR NUKLIR BEKAS REAKTOR RSG-GAS PADA RAK BERBAHAN ALUMINIUM ]]>
<![CDATA[Pungky Ayu Artiani]]>;
<![CDATA[Mirawaty Mirawaty]]>;
<![CDATA[Kuat Heriyanto]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.2.3280
<![CDATA[The use of stainless steel in nuclear spent fuels storage racks of interim storage facility potentially causes galvanic corrosion so material replacement of spent fuels storage racks needs to be considered. Galvanic corrosion is due to interaction between aluminum as in nuclear fuel of Multipurpose Research Reactor - G. A. Siwabessy (RSG-GAS reactor) and stainless steel as storage rack material for nuclear spent fuel. The purpose of this study is to assess the criticality of nuclear spent fuel storage racks using aluminum material to reduce galvanic corrosion effect. criticality guarantee is necessary to maintain the safety of nuclear spent fuel interim storage facility. Optimum criticality of aluminum storage rack is assessed by varying pitch size and calculating neutron absorption rate performed under normal condition (not an accident). Criticality (keff) calculation is performed using Monte Carlo N-Particle version 6 (MCNP6) code. The model used in this study is a 3-dimensional model where racks are filled fully with spent fuels in the storage pool. The calculations show that nuclear spent fuel storage rack stainless steel materials (keff = 0.6810) compared to aluminum (keff = 0.7709) in the size of 127 mm pitch, keff increased by 13.20%, but still within the value of safety requirement (keff ≤ 0.95). keff is affected by pitch size, where the reduction in the size of 1 mm pitch can increase the keffof 14.24%. Neutron absorption rate also affects keff, where neutron absorption rate in aluminum storage racks smaller than stainless steel. Storage racks using aluminum materials meet safety aspect to use as a storage rack in nuclear spent fuel interim storage facility because it has a keff less than 0.95 on a pitch size more than 112 mm (keff = 0.9196).]]>
<![CDATA[DIP-COATING PROCESS OF ZIRCALOY-2 FUEL CLADDING WITH COLLOIDAL GRAPHITE ]]>
<![CDATA[Etty Mutiara]]>;
<![CDATA[Meniek Rachmawati]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.1.3148
<![CDATA[The intensive research on high discharge burn-up of Light Water Reactor (LWR) fuel element have been continuously performed due to the extension of fuel element’s utility life. One of this research was allowing for alteration of the existing zirconium-based clad system through coating. A coating technique with the coating layer thickness of 10 - 30 µm will improve the corrosion resistance of cladding without changing the dimension of cladding. The scope of this current research is to obtain the zircaloy-2 cladding coated with ZrC layer by dipping process of zircaloy-2 specimens in colloidal graphite at room temperature. The dip-coated specimens undergo heating process at 700 oC, 900 oC and 1100 oC respectively in Argon gas atmosphere for 1 hour and are subsequently characterized by optical microscope and XRD. The optical microscope images show that the coating layers thickness are increased as the heating temperature increased. The coating layers thickness are 10 μm, 20 - 40 μm and 100 μm for the specimens heated at 700 oC, 900 oC and 1100 oC respectively. The calculated diffusivity of carbon into zircaloy-2 cladding for the coated specimens at 700 oC, 900 oC and 1100 oC are 3.10216E-11 cm2s-1; 3.60479E-11 cm2s-1 and 4.00613E-11 cm2s-1 respectively. From XRD examination analysis reveals that the ZrC phase appears in the specimens heated at 1100 oC but it is not the case for specimens heated at both 700 oC and 900 oC. The coating layer of specimens heated at both 700 oC and 900 oC mostly consists of carbon. At these heating temperatures, carbon atoms have diffused into zircaloy-2 and substituted the zirconium atoms with a limited occupation to form C-Zr solid solution. At the temperature of 1100 oC, due to the increase in vibration energy, the carbon atoms have enough energy to diffuse to form the carbide phase. Heating process at higher than 700 oC, however, will degrade the zircaloy-2 cladding. It is concluded that the dip-coating process of zircaloy-2 cladding in graphite colloid with subsequent high temperature heating is not the proper method to gain the ZrC-coated zircaloy-2 cladding. Therefore, as for the future works of this research, the others method should be searched and investigated to obtain the proper ZrC coating process on LWR zircaloy cladding which fulfills the dimension and quality requirements.]]>
<![CDATA[PERHITUNGAN BURN UP PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI POTONGAN BAGIAN MIDDLE ]]>
<![CDATA[Aslina Br. Ginting]]>;
<![CDATA[Supardjo Supardjo]]>;
<![CDATA[Yanlinastuti Yanlinastuti]]>;
<![CDATA[Sutri Indaryati]]>;
<![CDATA[Boybul Boybul]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 2 (2020): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2020.26.2.5864
<![CDATA[BURN UP CALCULATION OF THE MIDDLE PART OF IRRADIATED U3Si2/Al FUEL PLATE WITH 4.8 gU/cm3 DENSITY. Post-irradiation examination with non-destructive and destructive methods is required to study the performance of U3Si2/Al fuel during irradiation at the G. A. Siwabessy Multipurpose Reactor (RSG-GAS). The purpose of the burn-up calculation is to discover the 235U content that burned in a fuel plate of U3Si2/Al during use in RSG-GAS and to prove the burn-up obtained using destructive method by calculation using Origen Code. Therefore, this research performed a burn-up calculation of U3Si2/Al fuel plate (with a density of 4.8 gU/cm3) using destructive method through physicochemical analysis. The physicochemical analysis was aimed to determine the isotope content of 235U, 134Cs, and 137Cs in the irradiated U3Si2/Al fuel plate. The fuel plate was cut at the middle using a diamond cutting machine in a hot cell to obtain a sample cut with a weight of M-1 = 0.056 g, and a duplicate cut was also prepared with a weight of M-2 = 0.075 g. The sample was then dissolved using HCl 6 N and HNO3 6N. The results of dissolution was UO2(NO3)2 solution containing isotopic fission such as cesium and uranium. The solution of UO2(NO3)2 was pipetted at 1 mL and then transferred from the hot cell to R.135 (outside the hot cell) to separate 134Cs and 137Cs from 235U by a cation exchange method using zeolite from Lampung. The solution of UO2(NO3)2 was pipetted at 100 µL into a vial in duplicates and then 1000 mg zeolite was added. The cation exchange process was then carried out by stirring for 1 h. The separation process resulted in 134Cs and 137Cs isotopes in the solid phase, while uranium isotopes (238U, 235U, 234U), plutonium (239Pu, 238Pu) and other isotopes were in the liquid phase. Separation of uranium from other isotopes was done by anion exchange column method using 1200 mg of Dowex resin. The isotope content of 235U, 134Cs, 137Cs, 239Pu, and 238Pu was then measured using α/- spectrometer, the values of which were subsequently used for burn-up calculation. The result of the burn up calculation is 41.6565 ± 1.687%. The value of burn-up is not significantly different from the burn-up obtained from the calculation using Origen Code, which is 40%.]]>
<![CDATA[EFFECT OF CURRENT, TIME, TYPE OF FEED AND CATHODE ON ELECTROPLATING PROCESS OF URANIUM SOLUTION ]]>
<![CDATA[Sigit]]>;
<![CDATA[Ghoib Widodo]]>;
<![CDATA[Bangun Wasito]]>;
<![CDATA[Kris Tri Basuki]]>;
<![CDATA[Fahrunissa]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.1.3155
<![CDATA[Electroplating of uranyl nitrate and process effluent has been carried out in order to collect uranium contained therein using Pt/Pt and Pt/SS electrodes at various current and time values. The materials used as electrode were Pt (platinum) and SS (Stainless Steel). Feed solution of 250 mL was entered into a beaker glass equipped with Pt anode - Pt cathode or Pt anode - SS cathode and connected to direct current source from DC power supply at various current values and for various time periods so that uranium precipitated and deposited onto the cathode. Upon deposition completion, the cathode was removed and weighed to determine weight of precipitates, while the solution was analyzed to determine uranium concentration decrease after electroplating process. The experiment shows that a relatively good time to acquire uranium deposit at the cathode was 1 hour at 7 amperes, with uranyl nitrate as feed and Pt (platinum) as cathode. In these conditions, uranium deposit attached to the cathode amounted to 74.96% of the original weight of uranium oxide in the feed or 206.5 mg by weight. The use of Pt cathode for uranyl nitrate feed, and SS and Pt cathodes for effluent feed resulted in uranium specific weight at the cathode of 12.99 mg/cm2, 2.4 mg/cm2 and 5.37 mg/cm2 respectively for 7 Ampere current for 1 hour electroplating time.]]>
<![CDATA[PROSES EKSTRAKSI-STRIPPING UO2(NO3)2 BERIMPURITAS HASIL PELARUTAN DARI YELLOW CAKE ]]>
<![CDATA[Anwar Muchsin]]>;
<![CDATA[Ghoib Widodo]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.1.3172
<![CDATA[Uranyl nitrate/UO2(NO3)2 is a basic material/bait that is used as fuel both a research reactor fuel and power reactors. The process of extraction-stripping necessary to UO2 (NO3)2, because in the UO2(NO3)2 is still contained impurities are entrained during the dissolution process yellow cake. To obtain a solution of UO2 (NO3)2 which meets the requirements of pure nuclear (nuclear grade), it should be a process of purification by extraction-stripping using three (3) parameters of the feed flow rate, the acidity of the feed to the extraction process, and acidity in the stripping process. The mechanism of the extraction process takes UO2 (NO3)2 will go into the organic phase TBP, whereas impurity are in the water phase (refined). Uranium in the organic phase stripping process is carried out by determining the acidity (as a parameter), in order to obtain a mixed solution UO2 (NO3)2 pure nuclear (nuclear grade). As a bait used uranyl nitrate solution with levels of uranium amounted to 225.6282 gU/L. The results extraction process obtained the rate of feed water 15 L/h, the acidity of the feed to the extraction process by 3 M and obtained levels of uranium amounted to 48.5365 gU/L, while the stripping process is obtained UO2 (NO3)2 amounted to 64.7860 gU/L the acidity of 0.04 M.]]>
<![CDATA[ANALISIS KUAT SUMBER NEUTRON DAN PERHITUNGAN LAJU DOSIS NEUTRON TERAS AWAL RDE ]]>
<![CDATA[Suwoto]]>;
<![CDATA[Hery Adrial]]>;
<![CDATA[Zuhair]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.1.3119
<![CDATA[The reactor core RDE (Experimental Power Reactor) has non-annular cylindrical shape with coated fuel particles TRISO kernel in the spherical form (pebble) and cooled by helium gas. The RDE reactor core design is very safe, because it adopts the high-temperature reactor technology (HTGR) with inherent passive safety. The designed input helium temperature of RDE core is about 250 °C and the output of helium gas temperature helium of the reactor core of 700 °C. Beside of generating electricity, the RDE reactor is designed to generate high temperature gas that can be used for any purposes such as cogeneration (heat process application research). The spherical fuel called as pebble containing thousand of kernel TRISO-coated fuel particles of uranium dioxide 17 % (235U) enriched. TRISO coating comprises four layers, namely: porous carbon buffer layer, inner part pyrolytic carbon layer (IPyC, Inner Pyrolitic Carbon), silicon carbide layer (SiC) and a layer of pyrolytic carbon outer part (OPyC, Outer Pyrolitic Carbon). Analysis of neutron source strength and preliminary calculation of neutron dose rate on the RDE core were performed using Monte Carlo MCNP5v1.2 code. Double heterogeneity treatment in TRISO-coated fuel particles kernel and the pebble of the core are applied. Utilization of EGS99304 code to determine 640 amount of energy group structures (SAND-II neutron group structures) are used in the neutron spectrum calculation in the reactor. The reactor core is divided into 100 zones (10 in radial and 10 in axial directions). An analyses calculation result shows that the neutron source strength of the core is about 8.47027X1017neutron/sec. The neutron dose rate distributions are determined using a conversion factor of flux-to-dose taken from International Commission on Radiological Protection (ICRP) and National Council on Radiation Protection and Measurements (NCRP). The preliminary calculations result of neutrons dose rate using ICRP-21 and NCRP-38 conversion factor for radiation workers in the radial direction in the shield biological are already weakened as low as 6.69915 and 6.9964 μSv/h, respectively, at a position of 215 cm from the center part of the core. This result analysis indicate that radiation workers are safe and protected from neutron radiation sources in accordance with the regulatory requirements. Overall, from the calculation and analysis, it appears that the radiation and biological shielding model meets to the standards of radiation safety required. The results of this study can be used to complete the design data thickness of shielding of the reactor.]]>
<![CDATA[ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE ]]>
<![CDATA[Pande Made Udiyani]]>;
<![CDATA[Sri Kuntjoro]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.1.3160
<![CDATA[Experimental power reactor (RDE), which is planned to be constructed by BATAN, is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products to be remained in the core. To ensure the safety for workers and public surroundings of the RDE site and to meet the regulatory body requirements for construction and operation, research is needed for environmental radiation in normal and abnormal condition. This study is focused on normal operating condition, while for the accident condition study will be carried out after the final design has been completed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source terms released into the environment based on postulation of normal or routine operation of the RDE-10 MWth. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitations during the fabrication. Mechanisms of fission products release from the fuel to the environment were created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated by the use of ORIGEN-2 whose library has been modified for HTGR type, while assumptions on TRISO fuel defects and release fraction for each compartment of the RDE safety systems are defined by a reference parameter. The results show that the important source terms of RDE are group of noble gases (Kr and Xe), halogen (I), Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTR-10 source terms with the same power.]]>
<![CDATA[PENENTUAN KEMURNIAN RADIOKIMIA 99mTc-MIBI SECARA CEPAT DAN PRAKTIS MENGGUNAKAN DOSE CALIBRATOR ]]>
<![CDATA[Maskur Maskur]]>;
<![CDATA[Enny Lestari]]>;
<![CDATA[Endang Sarmini]]>;
<![CDATA[Yayan Tahyan]]>;
<![CDATA[Amal Rezka Putra]]>;
<![CDATA[Dede Kurniasih]]>;
<![CDATA[Adang Hardi Gunawan]]>
<![CDATA[ Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 1 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir ]]>
Publisher : <![CDATA[Badan Riset dan Inovasi Nasional ]]>
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DOI: 10.17146/urania.2017.23.1.3195
<![CDATA[Generally, the determination of radiochemical purity begins with Radiochemical separation using thin layer chromatography (TLC), and then measurement using a gamma counter, but this method cannot be applied in hospitals since there is only a dose calibrator available in the hospitals. Therefore, we conducted such an investigation using an alternative method which is more practical and faster, ie begins with radiochemical separation using extraction method, and then measurement using a dose calibrator. The determination was carried out in which 99mTc-MIBI was extracted using a mixture of saline and chloroform with volume ratio of 1:1, the mixture was then extracted. The 99mTc - MIBI was in organic phase (chloroform), whereas 99mTcO2 and 99mTcO4- were both in water phase (saline). Then, each phase of the sample was taken with the same volume, and then the radioactivity was measured using a gamma ionization chamber and the radiochemical purity was calculated. Radiochemical purity with extraction process method Compared measuring using standard methods of TLC Furthermore, the results of the extraction were compared with the standard method of TDC. The results showed that the radiochemical purity of 99mTc-MIBI using the two methods were almost the same, ie =98.34 % ± 0, 65 % for TLC standard method and = 97.15 % ± 0,56 % for extraction method with good measurement precision. The time of determination for extraction method was 20 minutes and 180 minutes for TLC standard method respectively. These results indicated that the extraction method can be used as an alternative method for determination of radiochemical purity of 99mTc-MIBI radiopharmaceutical. The advantage of extraction method that it is faster compared to the TLC Method.]]>
