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PENENTUAN RECOVERY NEODYMIUM DAN CERIUM PADA PROSES PEMISAHAN ISOTOP DARI SIMULASI PEB U3Si2/Al PRA IRADIASI Noviarty Noviarty; Erlina Noerpitasari; Rosika Kriswarini; Agus Jamaludin; Sutri Indaryati; Sayyidatun Nisa; Samsul Fatimah
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 26, No 1 (2020): Februari, 2020
Publisher : website

PENENTUAN RECOVERY NEODYMIUM DAN CERIUM PADA PROSES PEMISAHAN ISOTOP DARI SIMULASI PEB U3SI2/Al PRA IRADIASI. Neodymium (Nd) merupakan salah satu isotop hasil fisi yang digunakan sebagai indikator burnup bahan bakar nuklir karena isotop Nd merupakan hasil fisi yang stabil, mempunyai yield yang cukup besar, tidak mudah menguap dan mempunyai tampang lintang penangkap neutron yang rendah. Penentuan burnup menggunakan isotop Nd ini dapat mendukung dan memverifikasi hasil penentuan burnup yang telah dilakukan sebelumnya yaitu menggunakan isotop Cs, U dan Pu. Penentuan burnup menggunakan isotop Nd dilakukan secara merusak (Destructive Test, DT) berdasarkan kepada penentuan isotop spesifik yang dilakukan secara fisikokima menggunakan peralatan spektrometer massa. Berdasarkan metode ASTM E 321-69 untuk melakukan analisis isotop Nd diperlukan pemisahan Nd dari Ce karena Isotop 142Ce mengganggu pengukuran 142Nd yang disebabkan oleh sifat isobar kedua isotop tersebut. Fenomena ini akan mempengaruhi hasil perhitungan burnup.. Selain hal tersebut, keberadaaan isotop Nd dalam bahan bakar PEB U3Si2/AL masih bercampur dengan isotop hasil fisi lainnya seperti isotop 235U, 239Pu, 148Nd, 137Cs, 152Eu, 90Sr,143Ce,103Ru, 95Zr, dan 95Nb. Oleh karena itu untuk memperoleh nilai yang akurat sebagai indikator burnup maka isotop-isotop tersebut harus dipisahkan. Sebelum dilakukan pemisahan isotop Nd dan Ce maka dilakukan beberapa tahap proses pemisahan isotop Cs (137Cs,134Cs), isotop U (234U, 235U,236U,238U), isotop Pu (239Pu, 238Pu) dan isotop-isotop lainnya (152Eu, 90Sr, 103Ru, 95Zr, dan 95Nb). Penentuan recovery ini bertujuan untuk mengetahui jumlah Nd dan Ce pada setiap tahapan proses pemisahan terhadap isotop-isotop lain, sebelum sampai pada tahap pemisahan Nd dan Ce. Penentuan recovery dilakukan secara simulasi terhadap larutan PEB U3Si2/Al pra iradiasi yang ditambah larutan standar Cs, U, Nd dan Ce. Hasil recovery Nd dan Ce yang diperoleh dari tahapan proses pemisahan Cs menggunakan zeolit masing-masing sebesar 69,91% dan 64,42%, sedangkan hasil recovery Nd dan Ce pada tahapan proses pemisahan Pu dan U diperoleh sangat baik yaitu mendekati nilai 100%. Recovery Nd dan Ce ini selanjutnya digunakan sebagai koreksi pada perhitungan pemisahan Nd dan Ce dari larutan PEB, yang akan menjadi acuan dalam penentuan burnup bahan bakar PEB U3Si2/Al menggunakan indikator Nd.Kata kunci: Nd recovery, Ce recovery, indikator burnup, PEB U3Si2/Al

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): Oktober, 2020
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6071

OPTIMASI PROSES ELEKTRODEPOSISI UNTUK PENGUKURAN ISOTOP 242Pu DENGAN SPEKTROMETER ALPHA. Dalam penelitian ini digunakan metode elektrodeposisi untuk penyiapan sumber isotop 242Pu untuk pengukuran dengan spektrometer alpha. Tujuan penelitian ini adalah untuk mendapatkan parameter optimal proses elektrodeposisi sehingga diperoleh sumber alpha isotop 242Pu dengan spektrum yang baik dan hasil maksimal dari proses elektrodeposisi. Sampel standar isotop 242Pu dengan aktivitas tertentu dibuat dari larutan standar plutonium. Percobaan dilakukan untuk menetapkan parameter yang memiliki pengaruh signifikan terhadap efisiensi proses elektrodeposisi. Parameter proses elektrodeposisi yang dipelajari antara lain pengaruh waktu, arus listrik dan jarak anoda katoda menggunakan larutan elektrolit ammonium sulfat pH 3,5. Hasil optimasi proses elektrodeposisi diperoleh waktu optimum 2,5 jam, arus listrik 1,4 A dan jarak antara anoda katoda 10 mm. Hal ini menunjukkan bahwa parameter tersebut merupakan kondisi terbaik untuk deposisi isotop 242Pu. Hasil pengukuran dengan spektrometri alpha menunjukkan bahwa resolusi spektrum yang baik untuk sumber isotop 242Pu, dengan kedapatulangan proses elektrodeposisi sebesar 95,25%, dengan presisi sebesar 2,82%.Kata kunci: Elektrodeposisi, isotop 242Pu, kuat arus, waktu, spektrometer alpha

ANALISIS ISOTOP 137Cs DALAM PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI DENGAN METODE PENGENDAPAN CHLOROPLATINATE DAN KOLOM PENUKAR KATION RESIN DOWEX Arif Nugroho; Boybul Boybul; Sutri Indaryati; Iis Haryati; Rosika Kriswarini; Erlina Noerpitasari
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 26, No 3 (2020): Oktober, 2020
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6055

PEMISAHAN CESIUM DALAM PEB U3Si2/Al DENSITAS 4,8 gU/cm3 PASCA IRADIASI MENGGUNAKAN METODE PENGENDAPAN CHLOROPLATINATE DAN KOLOM PENUKAR KATION RESIN DOWEX. Pelarutan PEB U3Si2/Al dilakukan di dalam hot cell 109 menggunakan larutan HCl 6 M dan HNO3 6 M. Pemisahan cesium dengan uranium dalam larutan uranil nitrat dilakukan menggunakan metode pengendapan chloroplatinate dan kolom penukar kation. Pada metode pengendapan chloroplatinate, terlebih dahulu dilakukan proses penghilangan unsur-unsur yang akan mengganggu dalam proses pengendapan, selanjutnya dilakukan pemisahan heavy element seperti uranium atau plutonium dengan cara mengendapkan isotop 137Cs dalam bentuk Cs2PtCl6. Kandungan isotop 137Cs di dalam endapan Cs2PtCl6 diukur menggunakan menggunakan spektrometer-γ. Pada metode kolom penukar kation resin Dowex, larutan uranil nitrat dipipet sebanyak 150 μL dan ditambah 2 mL aquadest kemudian ditambahkan Cs carrier sebanyak 20 μL dan 1 mL HCl 12 M. Campuran larutan tersebut digunakan sebagai umpan dimasukkan ke dalam kolom penukar anion yang berisi resin R-Cl-. Efluen yang keluar dari kolom penukar anion dimasukkan ke dalam kolom penukar kation yang berisi resin R-NH4+. Isotop 137Cs yang terikat resin R-NH4+ di dalam kolom kemudian dielusi menggunakan HCI 1 M. Efluen kemudian dikisatkan sampai diperoleh volume ± 2 mL. Kandungan isotop 137Cs dalam PEB U3Si2/Al densitas 4,8 gU/cm3 hasil pemisahan menggunakan metode chloroplatinate untuk potongan Middle kode M diperoleh rerata sebesar 0,000557 g/gPEB dan 0,000652 g/gPEB untuk potongan Bottom kode B, sedangkan dengan menggunakan metode kolom penukar kation resin Dowex diperoleh sebesar 0,000747 g/gPEB untuk potongan Middle kode M, dan 0,000934 g/gPEB untuk potongan Bottom kode B. Rekoveri pemisahan 137Cs dengan metode kolom penukar anion lebih besar dibandingkan dengan metode pengendapan Cs2PtCl6.Kata kunci: U3Si2/Al, chloroplatinate, resin Dowex, 137Cs

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.

ANALISIS KOROSI PIPA PENDINGIN SEKUNDER RSG-GAS DENGAN TEKNIK ELECTROCHEMICAL IMPEDANCE SPECTROSCOPY (EIS) Maman Kartaman Ajiriyanto; Rosika Kriswarini; Yanlinastuti Yanlinastuti; Diyah Erlina Lestari
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.2018.24.2.4421

Multi-purpose reactor G.A. Siwabessy has a secondary cooling system facility to extract heat generated from uranium fission reactions in the reactor. The cooling pipe material is a type of carbon steel. Secondary cooling media is water supplied from Puspiptek. To extend the lifetime of the pipe, chemical compound ZnPO4 was put on the water-cooling media as a corrosion inhibitor. The corrosion analysis of secondary pipe was performed by corrosion testing using EIS and tafel method. The electrochemically corrosion test was performed with three electrode cells in water media, 0.05 M HCl solution and 3% NaCl solution. The aim is to obtain the optimum inhibitor concentration added in the secondary coolant and to determine the inhibition mechanical as well as the inhibitor efficiency. EIS testing was performed in the frequency range 100 kHz to 0.02 Hz in HCl 0,05M. Test results show changes in electrical parameters such as Rct, Cdl and Rs. The inhibitor concentration increases then Rct value increase. At inhibitor concentration of 90 ppm obtained the highest Rct value is 81,29 ohm. Efficiency of the inhibitor addition produces maximum value is 66,75% at concentration of 90 ppm. The corrosion rate in samples on medium of 3% NaCl without inhibitor, moss inhibitor and corrosion inhibitor were 3.975, 3.576 and 3.215 mpy respectively. The addition of a mixture of corrosion inhibitors and moss inhibitors can decrease the corrosion rate up to 1.586 mpy quantitatively. The addition of inhibitors using ordinary water may inhibit interface reactions on the surface of the sample as indicated by the higher the inhibitor concentration the higher the Rct value. In NaCl 3% medium also shows the same phenomenon relatively that the corrosion mechanism can be inhibited by increasing of Rct value as well as in this condition, Rs value is constant relatively. The dominant fact that the addition of a corrosion inhibitor could decrease the corrosion rate. The moss inhibitors have a primary function not to decrease the corrosion rate.

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.

ANALISIS KOROSI PADUAN ZIRLO-Mo DALAM MEDIA NaCl MENGGUNAKAN METODE POLARISASI Maman Kartaman Ajiriyanto; Dian Anggraini; 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.2017.23.3.3221

Zirlo-Mo is a zirconium type of alloy, which is developed as cladding material for light water reactor. As cladding material, the important parameter to be characterized is its corrosion resistance properties. The goal of this research is to study the corrosion resistance of zirlo-Mo in NaCl medium with varied concentration of 0.03; 0.3; and 3 % and compare it to the those of zircaloy-2 dan zircaloy-4 in the same medium. Samples were mounted into disk form and metallographically prepared, i.e. treatment with abrasive paper of 500 to 1200 grit SiC. The analysis of zirlo-Mo corrosion rate was performed by polarization method with a potentiostat. The advantage of this polarization resistance method is that it takes shorter testing time if compared to weight gain method, and it can be applied to analysis at low corrosion rate of below 0.1 mpy. Other tests were also done to support the corrosion rate analysis such as microhardness test with microhardness tester and microstructure examination with optical microscope. The tests results show that in general zirlo-Mo, zircaloy-2 and zircaloy-4 possess very low corrossion rate, i.e. below 1 mpy. The corrosion rate of zircaloy-2, zircaloy-4 and zirlo-Mo in NaCl 3 % medium are 0.006; 0.007 and 0.590 mpy respectively. The corrosion rate of zirlo-Mo is relatively the highest due to its lath martensite or accicular structure, which has high hardness value. The lath martensite structure and high hardness properties make zirlo-Mo highly susceptible to corrosion. Hardness test results show that the hardness value of zircaloy-2, zircaloy-4 and zirlo-Mo are 194.79; 227.95 and 360.05 HV respectively. The existence of solid solution of Mo and Zr-α as well as the martensite or acicular structure of zirlo-Mo result in high hardness value of the alloy.

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.

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