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OPTIMASI PARAMETER PEMISAHAN Zr DALAM PEB U3Si2/Al PRA IRADIASI MENGGUNAKAN METODE KROMATOGRAFI PENUKAR ANION Yanlinastuti Yanlinastuti; Noviarty Noviarty; Iis Haryati; Syamsul Fatimah; Boybul Boybul; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (2019): 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.2019.25.2.5508

Determination of optimum parameters in the process of zirconium separation from U3Si2/Al fresh fuel solution by anion exchanger chromatography and analysis with UV-Vis spectrophotometer has been carried out. Zirconium is one of isotopes that can be used in the determination of fuel burn-up. The purpose of this study is to obtain optimal parameters of zirconium separation in standard materials as well as U3Si2/Al fresh fuel solution by anion exchanger chromatography. Optimization of Zr separation parameters was carried out using Dowex 1x-8 resin with a variation in weight of 0.5; 1.0; 1.5; 2.0 and 2.5 g, acidity of H2SO4 at 1; 2; 3; 4; 5 and 6 M, standard feed concentrations of Zr at 50; 100; 150 and 200 ppm and flow rate of 3M H2SO4 elution at 0.1; 0.3; 0.5; 0.7 and 0.9 mL/minute. To the Zr effluent as the result of separation was added Arsenazo III 0.1% as complexing agent, and subsequent analysis with UV-Vis spectrophotometry was done at a wavelength of 665.3 nm. The measurement results show that the optimal parameters of Zr separation are the use of 1.5 g resin with resulting recovery of 70.79%, acidity of 3M H2SO4 with recovery of 83.94%, feed concentration of 150 ppm with recovery of 87.64% at a flow rate of 0.5 mL/minute. When those optimal parameters were used in Zr separation from U3Si2/Al fresh fuel solution, the recovery obtained was 96.23%. These optimized parameters can also be used to separate Zr from U3Si2/Al irradiated fuel solution.

PEMISAHAN DAN ANALISIS CESIUM DALAM PEB U3Si2/Al TMU 2,96 gU/cm3 PASCA IRADIASI DENGAN METODE PENGENDAPAN CHLOROPLATINATE Arif Nugroho; Boybul Boybul; Sutri Indaryati; Iis Haryati; Rosika Kriswarini; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (2019): 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.2019.25.2.5466

Fuel burnup is an important parameter for determining the quality and integrity of irradiated nuclear fuel. The most important parameters in the calculation of burnup is the separation process and 137Cs content analysis sampled from the irradiated plate fuel. This research has conducted separation and analysis of the 137Cs contained in irradiated U3Si2/Al plate by chloroplatinate precipitation method. This method requires that the sample has to be made free from elements that may interfere with the precipitation process. The separation is done by precipitation of 137Cs isotope in the form of Cs2PtCl6. The precipitation process uses CsNO3 as non-active carrier with varied weight of 0.0263; 0.0501; 0.0754; 0.1036 ; 0.2064; 0.3020, and 0.4017 g. The CsNO3 carrier is added in order to obtain more ammount of precipitate of Cs in the form of Cs2PtCl6 because CsNO3 has similar chemical properties with 137Cs in uranyl nitrate solution. The 137Cs isotope is coprecipitated in Cs2PtCl6, and γ-spectrometry is used in the determination of the content of 137Cs in Cs2PtCl6 precipitate. The separation results show that addition of 0,1036 g CsNO3 has recovered 0.000328 g/g-sample of 137Cs from 0.000465 g/g-sample in irradiated U3Si2/Al solution, or a recovery of 70.4753%. Recovery value by this method is still lower when compared to other 137Cs separation methods due to longer steps of separation process, which may cause many errors in the sample preparation process.

PEMBUATAN STANDAR CESIUM DARI LARUTAN PEB U3Si2/Al PASCA IRADIASI DENSITAS 2,96 gU/cm3 DENGAN METODE PENGENDAPAN CsClO4 Aslina Br. Ginting; Boybul Boybul; Arif Nugroho; Yanlinastuti Yanlinastuti; Noviarty Noviarty; Gatot Wurdiyanto; Hermawan Chandra
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 2 (2019): 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.2019.25.2.5457

A considerable amount of irradiated nuclear fuel solutions exists in the hot cell area of Radiometallurgy Instalation (IRM). These solutions were generated from sampling activities for burnup test. The solutions cannot be disposed due to the present of 137Cs isotope and other heavy elements that have a long half time, but the solutions can be used as a secondary source of 137Cs standard for analysis purposes. Post-irradiation fuel analysis is always conducted by a valid method and with a calibrated spectrometer-γ. The standard used in the calibration of spectrometer-γ energy or activity is 137Cs isotope. There has been some restriction in the use of spectrometer-γ because spectrometer-γ calibration cannot be performed routinely due to the unavailability of standard sources. The availability of 137Cs isotope for R&D is still dependent on abroad suppliers. Therefore, in this research, PTBBN and PTKMR aim to develop a secondary standard source of 137Cs isotope with 10330 ± 411 Bq activity. The standard source was obtained from a solution of 137Cs separated from U3Si2/Al irradiated fuel. The separation was done by precipitation method with CsNO3 and HClO4 as the carrier. The result of the separation is 137CsClO4 precipitate, dried and weighed for obtaining greater activity and measured with a spectrometer-γ. The dried precipitate 137CsClO4 was subsequently packed as closed standard source (shield source). This standard sourced was certified by PTKMR as secondary standard isotope of 137Cs and is ready to be used for energy calibration of spectrometer-γ.

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

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.

KARAKTERISASI PADUAN AlMgSi UNTUK KELONGSONG BAHAN BAKAR U3Si2/Al DENSITAS URANIUM 5,2 gU/cm3 Aslina Br. Ginting; Supardjo; Yanlinastuti Yanlinastuti; Dian Anggraini; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (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.2018.24.1.4016

Increase in uranium density from 2.96 gU/cm3 to 5.2 gU/cm3 of U3Si2/Al fuel requires a compatible cladding. High density fuel possesses high hardness value, and the use of AlMg2 as cladding material may cause dog bone during rolling process. In addition to dog bone phenomena, swelling can also occur during the use of nuclear fuel in the reactor due to rising in fission products and burn up. For this reason, substitution of cladding material should be determined to suit the high density U3Si2/Al fuel. This research deals with the characterization of AlMgSi as a candidate for AlMg2 substitution. The characterization includes thermal analysis with DTA (Differential Thermal Analysis) dan DSC (Differential Scanning Calorimetry), hardness test with macrohardness tester, microstructure examination with SEM (Scanning Electron Microscope), and corrosion analysis by heating at 150 °C for 77 hours in an autoclave. Analysis results show that both AlMgSi and AlMg2 share stable heat compatibility with U3Si2/Al up to 650 °C. AlMgSi cladding has a hardness value of 115 HVN, higher than that of AlMg2, which is 70.1 HVN. Microstructure examination shows that AlMgSi has better morphology of interface bonding than AlMg2. Corrosion rate analysis also indicates that AlMgSi has lower corrosion rate. The results of thermal, hardness, microstructure and corrosion characterization reveal that the U3Si2/Al fuel of 5.2 gU/cm3 suits better with AlMgSi than AlMg2.

OPTIMASI PARAMETER METODE PEMISAHAN RADIONUKLIDA 95Zr DALAM LARUTAN PELAT ELEMEN BAKAR U3Si2/Al TERIRADIASI Yanlinastuti Yanlinastuti; Noviarty Noviarty; Iis Haryati; Ariyanti Saputri; Boybul Boybul; Supardjo
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.2021.27.3.6535

Radionuclide of 95Zr is one of the γ-radiation emitter radionuclides which has a short half-life so that it can be used as an indicator of burn up. In the irradiated U3Si2/Al fuel element plate there are several radionuclides of fission products emitting α, β and γ and when measuring radiation using γ-spectrometry, these radionuclides may interfere with each other resulting in inaccurate measurement results. Therefore, it is necessary to separate 95Zr for better accuracy in the measurement of γ-radiation with a γ-spectrometer. The purpose of this study was to obtain the optimum parameters for the separation of 95Zr in U3Si2/Al fuel element plate solution by ion exchange column method using Dowex 1x-8Cl- resin of 100-200 mesh in diameter. The optimum parameters obtained were then used for the separation of 95Zr in a solution of irradiated U3Si2/Al fuel element. It was found that the optimum parameter of acidity of H2SO4 in the feed was a concentration of 0.5 M, and the acidity of H2SO4 for elution was 1.0 M with a recovery of 98.20%. The weight of the resin to bind 95Zr radionuclide was 2.0 g with a recovery of 78.76%. The required feed volume was 50 µL with a recovery of 96.90% and the optimized flow rate was 0.1 mL/minute with a recovery of 96.72%. The optimum values were used for the separation of the U3Si2/Al fuel element plate solution irradiated with the code of Bottom (B), Middle (M) and Top (T). The measurement results show that the mean recovery of separation was 95Zr for code B-1 = 75.078%; B-2= 81.401%; M1=76,850%; M-2=83.806%; T-1=84.433%; and T-2 = 81.728% with the acceptability of the CV repeatability value smaller than the CV Horwitz value.

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

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 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

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.

ANALISIS KANDUNGAN CESIUM DAN URANIUM DALAM BAHAN BAKAR U3Si2/Al PASCA IRADIASI Boybul Boybul; Yanlinastuti Yanlinastuti; Arif Nugroho; Dian Anggraini; 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.2017.23.2.3568

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.

PEMISAHAN DAN ANALISIS 137Cs DARI LARUTAN PELAT ELEMEN BAKAR U-7%Mo/Al Dian Anggraini; Noviarty Noviarty; Yanlinastuti Yanlinastuti; Aslina Br. Ginting; Rosika Kriswarini; Arif Nugroho; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 3 (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.2016.22.3.3183

The aim of this research is to get a valid method of separating cesium from U-7Mo/Al fuel plate solution through determination of parameter of method (accuracy, precision, and recovery). Precipitation method and cation exchange method that are used refer to standard ASTM 690-000 and research result of U3Si2/Al. Parameter method determination has been done by using 150 μL sample (U-7%Mo/Al fuel plate solution, 50 μL of standard solution in 2 mL of HCl 0,1 N. The sample solution was undergone precipitation process by using HClO4 concentrated and 225 mg of CsNO3 as carrier in temperature 0oC for an hour, while exchange cation process was done by using 400 mg of resin zeolite Lampung. The analysis of 137Cs in CsClO4 and 137Cs - zeolite was done by gamma spectrometer. Determination of recovery method was done by comparing count value of 137Cs before and after separation process. Recovery of precipitation method was obtained 95.56 % with accuracy and precision measurement of 0.375 % and 1.875 % respectively, while recovery of cation exchange method obtained 26.73 %. To sum up, the results show that precipitation method better than exchange cation method for separation 137Cs from U-7Mo/Al fuel plate solution.

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