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All Journal PIN Pengelolaan Instalasi Nuklir Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir
Torowati Torowati
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Energy Materials Science & Nanotechnology

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PROSES STRIPPING UNTUK PEMISAHAN URANIUM DARI SOLVEN TBP-HEKSANA Ngatijo Ngatijo; Banawa Sri Galuh Galuh; Rahmiati Rahmiati; Torowati Torowati
PIN Pengelolaan Instalasi Nuklir Vol 12, No 23 (2019): Oktober 2019
Publisher : PIN Pengelolaan Instalasi Nuklir

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Abstract

ABSTRAK Telah  dilakukan proses  stripping untuk  pemisahan  uranium dari  solven  TBP- heksana hasil ekstraksi uranium pada preparasi analisis unsur pengotor dalam uranium oksida. Proses stripping pada pemisahan uranium dari fasa organik ke fasa air, bertujuan agar uranium dapat diendapkan dan dikonversi kembali menjadi uranium oksida. Kegiatan ini dilakukan untuk mendapatkan larutan yang efektif untuk proses stripping,  sehingga dapat digunakan sebagai acuan  dalam  pengolahan  limbah  organik  yang  ditimbulkan  dari  preparasi  analisis  unsur pengotor dalam uranium oksida menggunakan alat AAS. Proses stripping dilakukan menggunakan larutan HNO3  0,01M, air panas 60 oC dan H2SO4  0,1M; 0,3M; 0,5M, perbandingan fasa organik dan fasa air (1:2), kecepatan pengadukan 500 rpm selama 30 menit, stripping   dilakukan   sampai   4   tingkat.   Proses   stripping   menggunakan   HNO3     0,01M menghasilkan  efisiensi  total  sebesar  38,66%  dan  menggunakan  air  panas  60oC  sebesar47,21%. Stripping menggunakan H2SO4 0,1M, 0,3M dan 0,5M diperoleh efisiensi total masing- masing sebesar 84,08%, 96,13% dan  99,08%. Larutan paling efektif untuk stripping uranium dari solven TBP-heksana adalah H2SO4 dengan konsentrasi 0,5M. Kata kunci - stripping, uranium, limbah organik TBP-heksana, H2SO4  ABSTRACT Stripping process to separate uranium from the TBP-hexane solvents results of uranium extraction  has been carried out in the preparation of impurities analysis in uranium oxide. Stripping process in the separation of uranium from organic solvent to aqueous phase has aim to precipitate and convert it to uranium oxide. The objective of this work is to obtain the effective stripper solution for stripping process, that it can be used as a reference for treatment of organic waste from preparation of impurities analysis in uranium oxide by AAS. Stripping process was conducted using HNO3  solution at 0.01M, hot water at 60oC and H2SO4 solution with concentration at 0.1M, 0.3M, 0.5M, the ratio of organic phase and aqueous phase was (1:2), stirring speed of 500 rpm for 30 minutes, and the stripping was carried out up to 4 stages. Stripping process with 0.01M HNO3 solution gave results total efficiency at 38.66% and with hot water 60oC result in 47.21% of total efficiency. The total efficiency of stripping with H2SO4 solution at 0.1M, 0.3M and 0.5M are 84.08%, 96.13% and 99.08% respectively. The most effective stripper solution for uranium stripping from TBP-hexane solvents is H2SO4 solution with a concentration of 0.5M. Keywords - stripping, uranium, TBP-hexane organic waste, H2SO4
PURIFICATION OF INDONESIAN NATURAL GRAPHITE AS CANDIDATE FOR NUCLEAR FUEL MATRIX BY ACID LEACHING METHOD: CHEMICAL CHARACTERIZATION Deni Mustika; Torowati Torowati; Arbi Dimyati; Sudirman Sudirman; Adel Fisli; I Made Joni; Ratih Langenati
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 26, No 3 (2020): Oktober, 2020
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Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2020.26.3.6059

Abstract

PURIFICATION OF INDONESIAN NATURAL GRAPHITE AS CANDIDATE FOR NUCLEAR FUEL MATRIX BY ACID LEACHING METHOD: CHEMICAL CHARACTERIZATION. Graphite matrix in Pebble Bed Reactor (PBR) – High Temperature Gas Cooled Reactor (HTGR) has an important role as heat transfer medium, neutron moderator and structural material to protect fuel. Thus, graphite matrix must fulfill chemical and physical characteristics for PBR-HTGR fuel. Indonesia has graphite sources in several regions that can potentially be purified. This research aimed to purify Indonesian natural graphite by several variation of acids and to perform chemical characterizations. Natural graphite from flotation process was purified by several variations of acid, i. e., hydrofluoric acid (HF), sulphuric acid + nitric acid (H2SO4 + HNO3) and hydrofluoric acid + hydrochloric acid + sulphuric acid (HF + HCl + H2SO4) and subsequently followed by chemical characterizations such as purity level, ash content, and boron quivalent. The highest purity was obtained in the purification process by HF with carbon content up to 99.52%; this purity level fulfills the specification of nuclear graphite (>99%). Ash content analysis shows a value in compliance with the specification requirement, i.e., < 100 ppm, and boron equivalent value also fulfills the specification value of < 1 ppm. It can be concluded from this study that the graphite purified by acid leaching with HF can be used as fuel matrix candidate but is qualified as low quality. Futher research is required to produce high quality nuclear graphite, particularly research in the minimization of the impurity by evaporation at temperatures over 950 oC to by far lower the ash content.Keywords:  Indonesian natural graphite, purification, nuclear fuel matrix, acid leaching, chemical characterization.
Co-Authors Adel Fisli Arbi Dimyati Banawa Sri Galuh Galuh Deni Mustika I Made Joni Ngatijo Ngatijo Rahmiati Rahmiati RATIH LANGENATI Sudirman Sudirman