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Sutri Indaryati
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Economics, Econometrics & Finance Engineering Industrial & Manufacturing Engineering Materials Science & Nanotechnology Social Sciences Other

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PEMISAHAN CESIUM DALAM PEB U3Si2/Al PASCA IRADIASI DENGAN METODE KOLOM PENUKAR KATION MENGGUNAKAN RESIN DOWEX Arif Nugroho; Dian Anggraini; Boybul Boybul; Sutri Indaryati; Rosika Kriswarini; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 2 (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.2018.24.2.4306

Abstract

The burn up calculations of nuclear fuel by destruction method can be determined from the analysis of 137Cs isotope, as fission product. For this reason, the separation of 137Cs isotopes from uranium in irradiated fuel element plates of U3Si2/Al (PEB) needs to be done. The separation of 137Cs was done by using cation exchange column method with Dowex resin with weight variation of 1; 2; 3; 4; 5 and 6 g. This study aims to determine the content of 137Cs in irradiated fuel element plates of U3Si2/Al (PEB) using the cation exchange column method with Dowex resin to be compared with the result of 137Cs separation by cation exchange column method using Lampung zeolite and CsClO4 precipitation method. The irradiated fuel element plates of U3Si2/Al (PEB) PEB U3Si2/Al with a density of 2.96 gU/cm3 from the middle part of RI-SIE2 code and side plate 20 IDA 0045 was taken with weight 0.1103 g and dissolved with HCl 6N and HNO3 6N to obtain uranil nitrate solution. The U3Si2/Al post irradiated solution was 150 μL was added with 2 mL aquadest and added with Cs carrier 20 μL and 1 mL HCl 12 M. The solution mixture was used as feed into the anion exchange column (first column) containing the R-Cl resin –with the wight variation about 1; 2; 3; 4; 5 and 6 g. The effluent from the anion exchange column was introduced into the cation exchange column (second column) which contained R-NH4+ resin with weight variation of 1; 2; 3; 4; 5 and 6 g. The bounded 137Cs isotope in R-NH4+ resin in the column was being eluted using 1 M HCl of 15 mL and being vaporized until the volume reach ± 2 mL. The quantity of the 137Cs isotope in the effluent was measured by using a gamma spectrometer with a counting time of 500 s. The 137Cs isotope content in irradiated fuel element plates of U3Si2/Al with the density 2.96 gU/cm3is 0.000445 g/gPEB and the parameter of separation 137Cs with cation exchange column method using Dowex resin weighing 2 g with recovery is 98.67%. The result of separation of 137Cs by cation exchange column method is not much different from the result of separation 137Cs by cation exchange method and CsClO4 precipitation method. The 137Cs obtained can then be used for burn up calculations.
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.
KARAKTERISASI TERMAL PEB U-7Mo/Al SEBAGAI BAHAN BAKAR REAKTOR RISET Sayyidatun Nisa; Ariyanti Saputri; Sutri Indaryati; Noviarty Noviarty; Supardjo; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (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.2021.27.2.6420

Abstract

UMo/Al-based fuel is a candidate for research reactor fuel and as an alternative to U3Si2/Al fuel. This is because the UMo/Al alloy has a higher uranium density of 16.4 g/cm3 compared to U3Si2/Al of 12.2 g/cm3. In addition, the UMo alloy has a low neutron absorption cross section and its reprocessing process is relatively easy. The U-7Mo/Al alloy is able to maintain the γ-U phase during irradiation process so that the stability of the fuel in the reactor is maintained. The U-7Mo alloy has been fabricated into a U-7Mo/Al mini fuel element plate. The purpose of this study was to determine the thermal characteristics of U-7Mo/Al as inputs to the research reactor fuel fabricator in the manufacture of UMo/Al-based fuel. Thermal characterization was carried out for uranium, molybdenum, UMo and the fuel element plate of U-7Mo/Al using Differential Scanning Calorimetry (DSC). The thermal characterization results show that the U-7Mo/Al fuel element plate is thermally stable up to 600 °C because it undergoes an endothermic thermochemical reaction at a temperature of 632.962 °C with DH = 144.318 J/g. At 656.186 °C, the Al matrix and AlMg2 cladding are fused in the U-7Mo/Al with DH = 144.318 J/g. At temperatures of 952.810 °C to 1106.050 °C, the molten Al matrix diffuses with UMo to form a metastable U(Mo,Al)X compound and heating to a temperature of 1211.760 °C forms a UAlX layer (UAl2, UAl3 and UAl4) with DH = –22,721 J/g.
Co-Authors Agus Jamaludin ARIF NUGROHO Arif Nugroho Ariyanti Saputri Aslina Br Ginting Aslina Br. Ginting ASLINA BR. GINTING Boybu Boybu BOYBUL BOYBUL Dian Anggraini Erlina Noerpitasari Hanifah Rifaatul Mahmudah Iis Haryati Iis Haryati Jan Setiawan Noviarty Noviarty Noviarty Noviarty Rahmiati Rahmiati Rosika Kriswarini Samsul Fatimah Sayyidatun Nisa Supardjo Supardjo Supardjo Yanlinastuti Yanlinastuti Yanlinastuti Yanlinastuti Yanlinastuti Yanlinastuti