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 <d002> 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.
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.
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.
Metode tidak langsung pengujian klorida dilakukan dengan mereaksikan klorida dengan perak berlebih, kemudian kadar perak sisa yang tidak bereaksi dihitung. Penelitian ini bertujuan untuk mengetahui kelayakan suatu metode melalui beberapa parameter, seperti linieritas, akurasi dan presisi, limit kuantitasi, dan limit deteksi instrumen pada sampel tanpa matriks (A), sampel UO2 dengan penambahan larutan standar Cl setelah ekstraksi (B) dan sampel UO2 dengan penambahan larutan standar Cl sebelum ekstraksi (C).Diperoleh nilai regresi linier pada sampel A sebesar 0,997, pada sampel B sebesar 0,996, dan pada sampel C sebesar 0,995. Nilai recovery yang didapat pada sampel A sebesar 102,804 %, pada sampel B sebesar 98,924 %, dan pada sampel C sebesar 98,096 %.Nilai simpangan baku relatif yang didapatkan pada sampel A sebesar 1,4 %, pada sampel C sebesar 0,2 %, dan pada sampel B sebesar 0,5 %. Nilai limit deteksi pada sampel dengan matriks yang didapat dengan metode ini, yaitu 0,0958 µg/g dan 0,1024 µg/g, sedangkan nilai limit kuantitasi pada sampel dengan matriks yaitu 0,3195 µg/g dan 0,3415 µg/g. Berdasarkan hasil penelitian, dapat dinyatakan bahwa metode tervalidasi dan layak diterapkan di Laboratorium Kendali Kualitas -Instalasi Elemen Bakar Eksperimental.
Indonesia has natural graphite reserved in Sulawesi, Sumatera and Kalimantan islands. The highest graphite content was observed in Sanggau Region, West Kalimantan. Indonesian natural graphite has the potential to be used as fuel matrix in Pebble Bed Reactor (PBR) type - High Temperature Gas Cooled Reactor (HTGR). To increase self-reliance of graphite supply and decrease graphite import for fuel matrix purposes, it is necessary to master the purification technology of graphite. Graphite matrix in Pebble Bed Reactor (PBR) fuel has an important role. It is not only used as neutron moderator and structural material to protect nuclear fuel but also as heat transfer media. Therefore, the graphite matrix must meet the physical and chemical criteria specified for PBR fuel. This paper focuses on current status of purification of Indonesian natural graphite as a candidate for nuclear fuel matrix using hydrometallurgy and pyrometallurgy. Acid and acid mixtures such as HF, HNO3+H2SO4 and HF+HCl+H2SO4 were used for the hydromet-allurgy purification process, while Arc Plasma System were used for pyrometallurgy methods. Characterization for the purified graphite includes physical and chemical properties. The result showed that Indonesian natural graphite obtained from hydrometallurgy was classified into low nuclear grade graphite but it still could be used as nuclear fuel matrix PBR - HTGR. The graphitization degree needs to be increased and the impurity content needs to be decreased for the purified Indonesian graphite using hydrometallurgy method. The Indonesian natural graphite obtained from preliminary pyrometallurgy methods has 96.78 % graphite content according to SEM (surface) observation and the graphite content needs to be increased into nuclear grade graphite.
THE POTENTIAL OF INDONESIAN GRAPHITE AS RDE FUEL MATRIX. The development plan of Ex- perimental Power Reactor (RDE) in Indonesia is non-commercial and leads to the technology type of Pebble Bed Reactor (PBR) - High Temperature Gas Cooled Reactor (HTGR). The fuel used for PBR reactors is kernel dispersed in spherical fuel elements. The matrix used in PBR nuclear fuel is graphite which functions as a neutron moderator, fuel protective material and heat conductor. Domestication of the domestic fuel matrix needs to be conducted to improve national independence. Therefore, it is necessary to do research on the potential of local graphite to be used as RDE fuel matrix. This study focused on the identification and characterization of local and commercial graphite. The results are compared with the literature, how far it is fulfilling nuclear grade graphite for PBR fuel matrix. Characterization of graphite includes phase analysis with XRD, micro- structure with SEM, surface area/porosity, impurities determination with AAS, ICP-OES and NAA, equivalent boron content, carbon content, density, particle size distribution and ash content. The characterization results show that the carbon content obtained was 87.0 ± 4.2% for local graphite and 100% for commercial graphite. Meanwhile, for the purposes of nuclear graphite it requires a carbon content of >99%. The impurity content in local and commercial graphite still does not meet the RDE fuel matrix standard. The results of XRD analysis show that the local graphite phase is the same as the commercial graphite phase, namely the 2H graphite hexagonal crystal system with the lattice group of P 63/mmc. Particle size distribution and surface area of local graphite are higher compared to nuclear graphite literature. The ash content of commercial graphite was 0.236 ± 0.029 and local graphite was 9.587 ± 0.010%. The results of this study indicate that the local graphite from the flotation still requires a further refinement process to obtain local graphite that can be used as a fuel matrix for RDE.
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.
ANALISIS KOMPOSISI UNSUR, DENSITAS, MAKROSTRUKTUR, DAN FASA PADUAN U-6Zr-xNb PASCA UJI KOROSI. Penelitian mengenai komposisi unsur, densitas, makrostruktur, dan fasa paduan U-6Zr-xNb pasca uji korosi telah dilakukan. Analisis komposisi paduan dilakukan sebelum uji korosi yang meliputi uji kadar uranium dengan titrasi potensiometri, uji kadar pengotor dengan Atomic Absorption Spectroscopy (AAS) serta uji kadar Zr dan Nb dengan X-Ray Fluorescence (XRF). Analisis komposisi paduan bertujuan untuk memastikan bahwa bahan bakar U-6Zr-xNb memenuhi syarat kualitas bahan bakar nuklir. Uji densitas dilakukan untuk menjadi salah satu parameter dalam menghitung laju korosi, sedangkan pengamatan makrostruktur dan fasa paduan dilakukan untuk mengetahui kerusakan atau produk korosi serta lapisan/fasa yang terbentuk setelah terjadi korosi. Hasil uji komposisi paduan U-6Zr-xNb dengan XRF maupun titrasi potensiometri menunjukkan bahwa kadar uranium sudah mendekati kadar yang syaratkan, sedangkan untuk uji pengotor dengan AAS menunjukkan adanya kadar pengotor melebihi yang disyaratkan untuk bahan bakar nuklir antara lain Al, Fe dan Si. Berdasarkan hasil tersebut, bahan bakar U-6Zr-xNb masih memenuhi persyaratan. Dilihat dari sifat neutroniknya, unsur Al, Fe dan Si memiliki tampang serapan neutron yang rendah. Hasil uji densitas sampel U-6Zr, U-6Zr-1Nb, U-6Zr-4Nb, dan U-6Zr-7Nb masing-masing sebesar 16,9798 g/mL, 16,6115 g/mL, 15,594 g/mL, dan 15,3564 g/mL. Pengamatan makrostruktur paduan pasca korosi menunjukkan adanya bercak hitam yang merupakan hasil oksidasi U (IV) menjadi U (VI). Paduan U-6Zr-xNb mengalami korosi paling besar pada media air bebas mineral. Kerusakan pada permukaan paduan semakin menurun seiring bertambahnya presentase berat Nb dalam paduan. Hasil karakterisasi paduan pasca korosi menggunakan XRD menunjukkan bahwa sampel U-6Zr dan U-6Zr-1Nb terbentuk fasa α, sedangkan untuk sampel U-6Zr-4Nb dan U-6Zr-7Nb terbentuk fasa γ. Lapisan oksida protektif Nb2O5 yang terbentuk sangat kecil, sehingga tidak terdeteksi oleh XRD.Kata kunci: Paduan UZrNb, komposisi, pengotor, korosi, densitas, makrostruktur, pembentukan fasa.
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