Search

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.

KARAKTERISASI KANDUNGAN URANIUM DAN UNSUR JEJAK PELET SINTER UO2 UNTUK FORENSIK NUKLIR. Forensik nuklir merupakan salah satu alat yang penting didalam keamanan nuklir terkait dengan penegakan hukum. Hal ini dikarenakan keberadaan bahan nuklir dan radioaktif memiliki potensi bahaya baik dari segi keselamatan, kesehatan hingga ancaman dalam keamanan nuklir. Didalam forensik nuklir, sidik jari adalah karakteristik bahan nuklir dan radioaktif untuk memberi informasi asal-usul suatu bahan nuklir sehingga diharapkan mempunyai data-data dari bahan nuklir dan radioaktif. Data-data diperoleh dari hasil karakterisasi berupa pengujian baik pengujian secara fisika ataupun kimia. Pengujian secara fisika seperti pengamatan visual, dimensi, fasa sedangkan secara kimia antara lain penentuan unsur bahan nuklir, penentuan konsentrasi unsur–unsur dalam bahan nuklir. Dalam makalah ini telah dilakukan pengujian kandungan uranium dan unsur jejak dalam bahan nuklir pelet uranium oksida (UO2) dengan tujuan untuk sidik jari dalam mendukung forensik nuklir yang ada di PTBBN, BATAN. Pengujian kandungan uranium dilakukan secara titrasi potensiometri sedangkan pengujian unsur jejak selain uranium dengan metode spektrofotometri serapan atom. Hasil rerata pengujian kandungan uranium dalam bahan nuklir dan radioaktif tersebut diperoleh antara 87% sampai 88% hal ini menginformasikan bahwa bahan tersebut adalah bahan nuklir UO2. Hasil pengujian kandungan unsur jejak selain uranium dalam pengujian ini berbeda pengayaan maka kandungan unsur jejaknya berbeda pula, sehingga dapat menginformasikan tentang tingkat pengayaan uranium yang dimiliki oleh pellet UO2 tersebut. Data-data tersebut dapat digunakan sebagai sidik jari dalam forensik nuklir sehingga dapat membantu penyidik dalam indentifikasi pada forensik nuklir apabila terjadi penyelewengan atau penyalahgunaan dari jenis bahan nuklir tersebut.Kata kunci: Uranium, pelet sinter, sidik jari, forensik nuklir.

Graphite material is extremely undissolvable to be turned into chemical solutions, therefore sample preparation is a serious problem faced in the determination of elemental impurity content in a graphite material. In this work, The nondestructive approach of instrumental neutron activation analysis (INAA) is applied to determine the concentration of multi-element in a graphite material, by employing both the forth floating process and the acid treatment method to the local Indonesian graphite. The sample was irradiated in the Rabbit system of G.A. Sywabessy Multi-Purpose Reactor at Serpong, Indonesia. The precision of the analysis was evaluated using certified reference materials which were obtained good performance with the most of concentration value in the range of 3 < zheta score < -3. Eleven elemental (Al, Sb, Co, Cu, La, Mn, Sc, Na, W, V, and Zn) concentration were determined in the forth floating process of the graphite. The Cu elemental is the most content with the value of 60,8 mg/kg or about 90% of total concentration content in graphite. Followed by the Sb content with a value of 5,5 mg/kg (about 8% of total impurities content in graphite). The remaining 2% includes the intermediate and the minor content of other impurity elements. After the acid treatment, the total concentration of impurities contained in the graphite material drastically decreases from 6.7% w/w to about 0,1; 0.6; and 0.59 % w/w for treatment employing the HF, HNO3+H2SO4,and HF+HCl+H2SO4 acid reagent, respectively. Cu element makes the largest contribution to reduce the concentration of impurities in graphite which decreased from 60,675 mg/kg to 1,088 mg/kg; 925 mg/kg and 835 mg/kg for HF, HNO3+H2SO4 and HF+HCl+H2SO4 acid reagent, respectively. In addition, Sb element concentration dropped dramatically from 5,514 mg/kg to 93 mg/kg using HF reagents. The other trace elements (As, Ba, Ca, Ce, Eu, Fe, Mg, Sm, and Th) were also identified in the acid reagent treated graphite sample which are suspected to derivates from the impurity reagent and or from contamination during the sample preparation. The treated HF for graphite was obtained the low purity grades approach for nuclear graphite.

Graphite matrix in Pebble Bed Reactor (PBR) fuel has an important role not only as neutron moderator and structural material to protect nuclear fuel, but also as heat transfer media. Therefore, the graphite matrix must meet the criteria of physical and chemical properties specified for PBR fuel. This paper focuses on the purification of the Indonesian natural graphite by using hydrometallurgy method with acid treatments. The characteristic of the purified graphite was studied for its specification compliance as a candidate of fuel matrix for PBR type of High Temperature Gas Cooled Reactor (HTGR). Acid and acid mixtures such as HF, HNO3+H2SO4 and HF+HCl+H2SO4 were used for the purification process. Crystal structure examination by X-Ray Diffraction indicates that the graphite sample was 2H poly type with hexagonal crystal structure and lattice group of P 63 m c space group. It was observed that the graphite sample purified by HNO3+H2SO4 mixture had the closest resemblance to single crystalline graphite with a deviation of 0.94 when compared to perfect graphite crystal. The density of graphite decreases from 2.3273 g/cm3 (before acid treatment) to 2.1808; 2.2203 and 2.2752 g/cm3 after treatment with HF, HNO3+H2SO4 and HF+HCl+H2SO4, respectively. These results are close to the theoretical density value of 2.26 g/cm3. The surface area decreases from 10.346 m2/g to 6.177; 5.831 and 7.63 m2/g for the treated graphite with HF, HNO3+H2SO4 and HF+HCl+H2SO4 respectively. However, these values are still higher than that of nuclear grade graphite (i.e. between 4.80 and 5.55 m2/g). The average diameter size of graphite decreased from 29.65 μm (before treated acid) into 23.12 μm (after treated acid). The Indonesian natural graphite obtained from acid purification treatment is potential to be used as matrix material for PBR - HTGR fuel, but further treatment is necessary.

PENENTUAN KADAR URANIUM DALAM SERBUK UO2 DARI YELLOW CAKE SECARA POTENSIOMETRI DAN GRAVIMETRI. Telah dilakukan penentuan kadar uranium dalam serbuk UO2 dari yellow cake yang berasal dari Pusat Teknologi Bahan Galian Nuklir (PTBGN). Serbuk UO2 yang ditentukan kadar uraniumnya merupakan hasil konversi yellow cake melalui jalur amonium diuranat (ADU) dan amonium uranil karbonat (AUK). Dalam penentuan kadar uranium diperlukan suatu metode yang valid sehingga diperoleh data yang akurat. Dalam kegiatan ini dibandingkan dua metode penentuan kadar uranium dalam serbuk UO2, yaitu secara potensiometri dan secara gravimetri. Tujuannya untuk menentukan mana metode yang memberikan hasil dengan ketelitian dan presisi tinggi. Penentuan kadar uranium secara potensiometri mengacu kepada ASTM C1267-11 yang dimodifikasi dengan mereduksi pemakaian pereaksi menjadi 10% dari metode asli, sedangkan metode gravimetri mengacu kepada ASTM C1453-00 (2011). Penentuan kadar uranium secara potensiometri berdasarkan oksidasi-reduksi, sedangkan secara gravimetri berdasarkan perubahan berat setelah serbuk UO2 dikalsinasi pada temperatur 900°C selama 3 jam hingga diperoleh berat konstan. Dari kegiatan ini diperoleh hasil rerata untuk penentuan kadar uranium dalam serbuk UO2 melalui jalur ADU dan AUK secara potensiometri masing-masing adalah (87,4019 ± 0,7873) % dan (87,5575 ± 0,4775) %, sedangkan secara gravimetri masing-masing adalah (87,5241 ± 0,0432) % dan (87,4492 ± 0,0440) %. Uji t dari kedua metode menunjukkan bahwa hasil penentuan kadar uranium tidak mempunyai perbedaan yang signifikan. Oleh sebab itu, penentuan kadar uranium dalam serbuk UO2 pada penelitian berikutnya digunakan metode gravimetri karena lebih mudah tahapan pengerjaannya dibandingkan dengan metode potensiometri.