Aslina Br. Ginting
Pusat Teknologi Bahan Bakar Nuklir – BATAN

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UJI TAK MERUSAK PELAT ELEMEN BAKAR U3SI2/Al DENSITAS URANIUM 4,8 gU/cm3 MENGGUNAKAN RADIOGRAFI SINAR-X DIGITAL Refa Artika; Rohmad Sigit; Helmi Fauzi Rahmatullah; Supardjo Supardjo; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 1 (2020): 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.2020.26.1.5822

Abstract

NON DESTRUCTIVE TEST FOR U3Si2/Al FUEL PLATE WITH 4,8 gU/cm3 URANIUM DENSITY USING DIGITAL X-RAY RADIOGRAPHY. Non-destructive test for irradiated U3Si2/Al fuel plate with 4,8 gU/cm3 uranium density and 20, 40, and 60% burn-up has been done. The test includes visual inspection, radiographic image analysis, and dimension measurement using digital x-ray radiography. The non-destructive test aimed to determine the performance, integrity, and reliability of U3Si2/Al fuel plate with 4,8 gU/cm3 uranium density during irradiation at RSG-GAS. Reliability and safety of the fuel during irradiation are indicated by the absence of anomalies in the post-irradiation examination result. Three samples of irradiated U3Si2/Al fuel element (FE) with 20, 40, and 60 % burn-up were transferred from RSG-GAS to RMI hot cell 101 through transfer channel of Interim Storage Spent Fuel (ISSF). The U3Si2/Al fuel plate were then transferred to hot cell 103 for non-destructive test using digital x-ray radiography with seven shots for good resolution radiographic images. The radiographic images were evaluated to obtain data of defects and dimensions such as meat width, fuel plate width, meat-cladding distance, right-left meat- cladding boundary distance at the far side (SJ) and the near side (SD). The X-ray radiography analysis shows the presence of white spot in the middle of U3Si2/Al fuel cladding with 20% burn- up. Meanwhile, indication of surface defects and other defects were not found for the 40 and 60% burn-up samples. The dimensional measurement data show no dimensional changes, which indicates solid integrity after irradiation in the RSG-GAS. The measurement data analisis can further be used for mapping the sampling spots for the next test.
PERHITUNGAN BURN UP PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI POTONGAN BAGIAN MIDDLE Aslina Br. Ginting; Supardjo Supardjo; Yanlinastuti Yanlinastuti; Sutri Indaryati; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 2 (2020): 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.2020.26.2.5864

Abstract

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%.
PENENTUAN BERAT ISOTOP CESIUM DAN URANIUM DALAM PELAT ELEMEN BAKAR U3Si2/Al DENSITAS 4,8 gU/cm3 Yanlinastut Yanlinastut; Sutri Indaryati; Boybul Boybul; Supardjo Supardjo; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 2 (2020): 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.2020.26.2.5874

Abstract

DETERMINATION OF WEIGHT OF CESIUM AND URANIUM ISOTOPES IN IRRADIATED U3Si2/Al FUEL PLATE WITH 4.8 gU/cm3 DENSITY AND 40% BURN UP. The fuel plate of U3Si2/Al with a density of 4.8 gU/cm3 was irradiated in the RSG-GAS multipurpose reactor with 40% burn- up. During irradiation, the amount of isotopes produced by fission reaction, such as cesium, uranium and other isotopes can be used for burn-up calculation to determine the fuel performance. Physicochemical analysis was performed to determine the weight of cesium and uranium isotopes in the irradiated fuel. The irradiated plate of U3Si2/Al was cut at the middle section (duplo) with a weight of M-1 = 0.065 g and M-2 = 0.075 g and dissolved in HCl 6N and HNO3 6N. The resulted was pipetted at 100 µL and then added into each of 2 mL solvent of HNO3, H2SO4 and HCl 0.1N. The separation of cesium was then carried out with the addition 1000 mg of zeolite from Lampung. Cesium isotope obtained from the separation was bound with zeolite in the solid phase while uranium was in the liquid phase. Cesium isotopes in the solid phase and in the liquid phase were measured by a spectrometer- for their radioactivity for 10000 seconds. The measurement was indicated by spectra of 134Cs at 604.7 and 795.8keV energy and 137Cs at 661.7keV. Meanwhile, the separation of uranium in the liquid phase was carried out using anion exchange column method with the addition of Dowex resin. The results of the separation were indicated by some spectra of U isotopes at various energies, 238U at 4,1943 MeV, 235U at 4,397 MeV, 236U at 4,494 MeV and 234U at 4,777 MeV. Weight calculation from the measurement data shows that the weight of 137Cs in 0.056 g of the irradiated U3Si2/Al fuel samples in HNO3, H2SO4 and HCl solvent are 4.08E-05 g; 4.53E-05 g ,and 4.53E-05 g respectively, while 134Cs weight are 4,84E-08 g; 5,30E-08 g and 5,20E-08 g respectively. Meanwhile, 137Cs isotope weight in 0.075 g fuel plate of U3Si2/Al are obtained by 5.93E-07 g; 5.90-05 g, and 5.88E-05 g, while the isotope weight of 134Cs in HNO3, H2SO4 and HCl solvent are 5,23E-07 g; 5,29E-07g and 5,008E-07g respectively. The isotopes weight of 238U, 235U, 236U and 234U in 0.056 g of fuel plale are 6.45E-03 g; 2.57E-03 g; 2.42E-06 g; and 1.90E-05 g, and the weight in 0.075 g fuel plale are 7.41E-03 g; 2.40E-03 g; 2.91E-06 g; and 2.17E-05 g. The isotopes weight of 134Cs, 137Cs, 238U, 235U, 236U and 234U in 0.056 g and 0.075 irradiated U3Si2/Al fuel plate obtained were relatively similarly. Data of isotopes weight are then used as input data for burn up calculation.
PROSES ELEKTRODEPOSISI UNTUK PENGUKURAN ISOTOP 242Pu DENGAN SPEKTROMETER ALPHA Boybul Boybul; Yanlinastuti Yanlinastuti; Arif Nugroho; Rosika Kriswarini; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 26 No. 3 (2020): 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.2020.26.3.6071

Abstract

OPTIMIZATION OF ELECTRODEPOSITION PROCESS FOR 242Pu MEASUREMENT BY ALPHA SPECTROMETRY. In this study electrodeposition method was used to design and set up an electrodeposition device for 242Pu source preparation for measurement by alpha spectrometry. The purpose of this study was to determine the optimal electrodeposition process parameters so as to obtain 242Pu alpha sources with good spectrum properties and maximum result from source preparation by electrodeposition process. Standard samples of 242Pu with specific activity were prepared from plutonium standard aqueous solution. Experiments were conducted to determine the main parameters that have significant influence on the deposition efficiency. The electrodeposition process parameters analyzed were the effects of time, electric current and anode-cathode distance using an electrolyte solution of ammonium sulphate at pH 3.5. The results of the optimization of the electrodeposition process were optimal time of 2.5 hours, electric current of 1.4A and cathode- anode distance of 10 mm. These parameters' values are the best conditions for the deposition of 242Pu. Alpha measurement results of the electrodeposited source of 242Pu also show good spectrawith electrodeposition process reproductibility of 95.3%.
INTERAKSI BAHAN BAKAR U3Si2-Al DENGAN KELONGSONG AlMg2 PADA ELEMEN BAKAR SILISIDA TMU 2,96 gU/cm3 PASCA IRADIASI Aslina Br. Ginting; Maman Kartaman Ajiriyanto; Supardjo Supardjo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 21 No. 1 (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.2015.21.1.2255

Abstract

INTERACTION OF U3Si2-Al FUEL ELEMENT WITH AlMg2 CLADDING ON POST IRRADIATION WITH LOADING OF URANIUM 2.96 gU/cm3. Interaction of U3Si2-Al fuel element with AlMg2 cladding on post irradiation of 2.96 gU/cm3 loading of uranium (TMU) of U3Si2-Al fuel elements plate (PEB) has been analyzed. The purpose of this research is to study the changes of microstructure of nuclear fuel elements during iradiation in reactor core. Understanding on interaction of U3Si2-Al fuel meat with AlMg2 cladding onpre and post irradiation needed to study the influence of radiation on fuel elements plate. PEB U3Si2-Al with 2.96 gU/cm3 by size 10 × 10 mm were heated in DTA (Differential Thermal Analysis) furnace with temperature variation at 450, 550, 650, 900 and 1350oC to perform pre irradiation test.Post irradiation samples were cut by size 2 × 10 mm as many as three samples taken from bottom, middle, and top of PEB in hotcell. The metallography preparation for each pieces of pre and post irradiation samples of U3Si2-Al fuel elements platewith 2.96 gU/cm3 weredone through steps mounting, grinding, polishing, and etching.Scanning Electron Microscope (SEM-EDS) were used to observe the pre irradiation microstructure of fuel elements U3Si2-Al with AlMg2 cladding interaction, while the post irradiationmicrostructure were observed by optical microscope in hot cell. The result show the interaction of U3Si2 with Al matrix or AlMg2 cladding in pre irradiation PEB U3Si2-Aloccurred agglomeration formed new compouds of U(Al,Si)x and UAlx formation. Agglomeration formation on heated pre irradiation samples were bigger while heating temperature increased. The post irradiation sampels shoed the oxide layer were formed outside the AlMg2 cladding and the inner side of caldding that contact to the fuel meat formedlight-grey U(Al,Si)xlayer at 1-3 micron of thickness.