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All Journal Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir Jurnal Sains Dasar Jurnal Fisika FLUX
Zuhair Zuhair
Center forNuclear Reactor Technology and Safety, National Nuclear Energy Agency, Puspiptek Area, Serpong Tangerang 15310, Indonesia
Earth & Planetary Sciences Education Materials Science & Nanotechnology Physics

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 1 
STUDI DAN KAJIAN DATA NUKLIR REAKTOR GENERASI-IV DENGAN SPEKTRUM NEUTRON CEPAT Suwoto, Suwoto; Zuhair, Zuhair
Jurnal Fisika FLUX Vol 10, No 1 (2013): Jurnal Fisika FLUX Edisi Februari 2013
Publisher : Lambung Mangkurat University Press

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.20527/flux.v10i1.2633

Abstract

Abstract. STUDY AND ASSESSMENT OF GENERATION IV REACTOR NUCLEARDATA WITH FAST NEUTRON SPECTRA. Generation IV International Forum (GIF)has evaluated and assessed NES of Gen- IV and selected six potential types ofreactors to be deployed in the next decade. Those include GFR, LFR, SFR, MSR,SCWR and VHTR. The first three reactors were fast neutron spectrum applied and therest reactors were thermal neutron spectrum used. The study and assessment focusedon the nuclear data characteristic parameter and nuclear data uncertainties of Gen-IVreactor with fast neutron spectrum. Until 2008, the accuracy target of nuclear datacross-sections used it in fast reactor spectrum calculation are relatively significantespecially for σ-capture, σ-fission, and σ-inelastic. Several differences of nuclear datacross-sections on minor actinide isotopes between expected and targeted parametersare observed such as σ-fission of Cm-244 isotope up to 10 times larger and σ-captureof 92-U-238 isotope around 1.5-2 times higher than targeted parameters. Uncertaintyand accuracy of minor actinide cross-sections for fast spectrum Gen-IV reactorsprovide relatively significant discrepancies (1.3 to 10 times higher) in term of accuracybetween expected and targeted parameters. Some differences of provided results fromany experimental and assessment data with several evaluated nuclear data files for Pbare found. Some discrepancies on integral parameter of fast spectrum Gen-IV reactorsbetween expected and targeted such keff, void reactivity and Doppler effects, peakpower and burn-up are clearly observed. Accurate and precise cross-sections data ofradiation captured and threshold reaction cross sections such as (n,2n), (n,3n), (n,p),(n,α) are necessary for fast reactors.Keywords: cross-sections, fast neutron spectrum, GFR, LFR, SFR, uncertainty, targetaccuracy
Studi Fleksibilitas dan Posibilitas Daur Bahan Bakar Reaktor Temperatur Tinggi (HTR) Adrial, Hery; Zuhair, Zuhair
Jurnal Fisika FLUX Vol 9, No 1 (2012): Jurnal Fisika Flux Edisi Februari 2012
Publisher : Lambung Mangkurat University Press

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.20527/flux.v9i1.3129

Abstract

Dewasa ini sejumlah institusi riset di dunia sedang mengembangkan teknologi reaktor temperatur tinggi (HTR) melalui berbagai program seperti PUMA, RAPHAEL, ANTARES, dll. Tujuan dari program ini adalah untuk mengembangkan HTR versi demontrasi dan komersial. HTR merupakan reaktor berpendingin gas temperatur tinggi, bermoderator grafit dengan spektrum neutron termal dan temperatur outlet teras hingga 1.000oC. Fasilitas energi ini dapat mencapai efisiensi termodinamika yang cukup tinggi (~80%) dengan kapabilitas generasi listrik dan produksi hidrogen. Makalah ini membahas fleksibilitas dan posibilitas daur bahan bakar HTR yang meliputi serangkaian daur bahan bakar komprehensif. Dikategorikan ke dalam 4 kelompok, yaitu daur bahan bakar uranium pengkayaan rendah (LEU), daur bahan bakar MOX, daur bahan bakar hanya plutonium dan daur bahan bakar berbasis thorium, daur bahan bakar HTR dievaluasi secara sistematis. Makalah ini juga mendiskusikan pertimbangan pemilihan daur bahan bakar untuk sejumlah HTR yang telah dioperasikan seperti AVR dan THTR Jerman, Peach Bottom dan Fort Saint Vrain USA, DRAGON Inggris, dll., dan yang sedang dalam proses desain seperti HTR AREVA Perancis, PBMR Afrika Selatan, dll. Hasil diskusi menyimpulkan bahwa daur LEU layak dipilih sebagai daur referensi untuk proyek HTR masa kini dan masa yang akan datang.
Studi Perhitungan Benchmark Kritikalitas Teras Metalik dan MOX di FCA Zuhair, Zuhair; Sembiring, Tagor M.; Yazid, Putranto Ilham
Jurnal Fisika FLUX Vol 8, No 2 (2011): Jurnal Fisika Flux Edisi Agustus 2011
Publisher : Lambung Mangkurat University Press

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.20527/flux.v8i2.3118

Abstract

The criticality experiments at FCA three cores have been done to obtainreliable original data as a benchmark test. The first two cores was mock-up of metallicfueled LMFBR’s and the other was a mock-up of a MOX fueled LMFBR. The criticalitycalculation of FCA cores was performed by using the Monte Carlo transport codeMCNP-4C in 2-D R-Z reactor geometry. The analysis was done with ENDF/B-VIcontinuous energy neutron cross-section library at room temperature. The MCNP-4Ccriticality prediction (keff) for metallic (XVI-1 and XVI-2) cores were underestimated in0.54% and 0.48%, respectively. The MCNP-4C criticality prediction (keff) for MOX(XVII-1) core showed best agreement with the experimental data where C/E value was0.99995. In general, it can be concluded that MCNP-4C calculations on FCA criticalitybenchmark experiments show a high accuracy for metallic as well as MOX cores.
Analisis Perhitungan Benchmark Keselamatan Kritikalitas Larutan Uranil Nitrat di Teras Slab 280T STACY Zuhair, Zuhair; Suwoto, Suwoto; Suharno, Suharno
Jurnal Fisika FLUX Vol 8, No 2 (2011): Jurnal Fisika Flux Edisi Agustus 2011
Publisher : Lambung Mangkurat University Press

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.20527/flux.v8i2.3115

Abstract

Criticality benchmark experiment at STACY critical facility is important forvalidation of computation technique and nuclear data library used in design of nuclearfuel cycle criticality safety. This paper discusses criticality safety benchmarkcalculation at STACY facility, which uses uranyl nitrate solution with MCNP-4C MonteCarlo transport code. The continuous energy nuclear data library was utilized inbenchmark calculation to complete criticality safety analysis. The MCNP-4C criticality(keff) prediction indicated overestimated results for all configurations except forconfiguration 131. The biases of calculation with criticality experiment (keff = 1) wereunder 0.26%. Configuration 140 calculation showed the most precisely agreement withC/E value of 1.0001. From these results, it can be concluded that the capability andreliability of MCNP-4C is constantly high in prediction of criticality accuracy for uranylnitrate solution at STACY 280T slab core.
Studi Sensitivitas Ketinggian Teras Reaktor dalam Desain Htr Pebble Bed Zuhair, Zuhair
Jurnal Fisika FLUX Vol 9, No 1 (2012): Jurnal Fisika Flux Edisi Februari 2012
Publisher : Lambung Mangkurat University Press

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.20527/flux.v9i1.3131

Abstract

HTR pebble bed adalah reaktor temperatur tinggi berbahan bakar pebble dan berpendingin gas helium dengan teras densitas daya rendah. Teras HTR pebble bed disebut teras grafit penuh karena mengggunakan struktur grafit sebagai moderator dan reflektor, partikel bahan bakar berlapis grafit dan elemen bakar grafit lengkap. Makalah ini mendiskusikan sensitivitas ketinggian teras reaktor dalam desain HTR pebble bed. Perhitungan dikerjakan dengan program transport Monte Carlo MCNP5 pada temperatur kamar. Berbagai opsi matriks bahan bakar UO2, PuO2 dan ThO2/UO2 dieksaminasi untuk konfigurasi teras dengan rasio F/M 1:0 dan F/M 1:1. Hasil perhitungan memperlihatkan kondisi kritis teras dengan bahan bakar PuO2 untuk rasio F/M 1:0 dicapai pada ketinggian 66 cm, ThO2/UO2 pada ketinggian 88 cm dan UO2 pada ketinggian 106 cm sedangkan ketinggian kritis konfigurasi teras PuO2 dengan rasio F/M 1:1 terjadi di 78 cm, ThO2/UO2 di 124 cm dan UO2 di 138 cm. Dari hasil ini dapat ditentukan ketinggian teras dengan konfigurasi rasio F/M 1:0 dan F/M 1:1 yang layak dipilih dalam desain HTR pebble bed dengan opsi bahan bakar UO2, PuO2 dan ThO2/UO2 guna mendapatkan reaktivitas yang dispesifikasikan. Dapat disimpulkan bahwa selain pengkayaan, radius kernel, fraksi packing partikel TRISO dan dimensi reflektor grafit, ketinggian teras reaktor merupakan salah satu parameter neutronik yang harus dipertimbangkan dalam desain HTR pebble bed dengan matriks bahan bakar yang spesifik.
STUDI PERHITUNGAN HTR PEBBLE-BED DENGAN BERBAGAI MODEL KISI KERNEL DAN KISI PEBBLE Zuhair, Zuhair
Jurnal Sains Dasar Vol 1, No 1 (2012): April 2012
Publisher : Faculty of Mathematics and Natural Science, Universitas Negeri Yogyakarta

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (400.55 KB) | DOI: 10.21831/jsd.v1i1.2349

Abstract

In pebble-bed reactor core, lattice models can be changed and even varies with the changing position of pebble. This sort of thing can be found mainly in the area near the vessel wall. Lattice model applied also depends on the procedure how the cylindrical vessel filled. Some lattice models such as SC, BCC, FCC, SH and HCP often used in the high temperature reactor to treat kernel randomness in the graphite matrix and pebble fuel in the reactor core. In this paper a series of calculations of the reactor multiplication factor (keff) conducted with various model of kernel and pebble lattices. The effect of lattice combination which implies on neutronics performance of HTR pebble-bed design is analyzed utilizing the Monte Carlo transport code MCNP5 and continuous energy nuclear data library ENDF/B-VI. MCNP5 calculations show consistency with the keff values which ​​are almost the same for all combinations of kernel and pebble lattices, but if observed further appears that the keff value is more dependent on pebble lattice than kernel lattice kernel. Kernel lattice provides only a less significant effect. The results of keff predictions of all lattice combinations conclude that whatever kernel lattice model utilized, the BCC pebble lattice model is better adopted in the calculation of HTR pebble-bed design with UO2, PuO2 and ThO2/UO2 fuel.
Analisis Kuat Sumber Neutron Dan Perhitungan Laju Dosis Neutron Teras Awal RDE Suwoto, Suwoto; Adrial, Hery; Zuhair, Zuhair
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 23, No 1 (2017): Februari 2017
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (487.916 KB) | DOI: 10.17146/urania.2017.23.1.3119

Abstract

Teras reaktor RDE (Reaktor Daya Eksperimental) berbentuk silinder non anular, mengadopsi teknologi HTGR (High Temperature Gas-cooled Reactor) berbahan bakar kernel partikel berlapis TRISO dalam bentuk bola (pebble) dan berpendingin gas helium. Desain teras reaktor RDE ini mengadopsi teknologi reaktor temperatur tinggi HTGR dengan keselamatan inherent pasif yang sangat aman. Temperatur keluaran panas gas helium teras reaktor RDE dirancang pada kisaran 700°C dengan temperatur masukan sekitar 250°C. Di samping menghasilkan listrik, reaktor RDE didisain menghasilkan panas temperatur tinggi yang dapat digunakan untuk keperluan kogenerasi lainnya (penelitian panas proses lainnya). Bahan bakar pada RDE berbentuk bola yang berisikan kernel partikel berlapis TRISO yang berupa uranium oksida (UO2) berpengkayaan 17%. Lapisan TRISO terdiri 4 lapisan yaitu lapisan karbon penyangga berpori, lapisan karbon pirolitik bagian dalam (IPyC, Inner Pyrolitic Carbon), lapisan Silikon Karbida (SiC) dan lapisan pirolitik karbon bagian luar (OPyC, OuterPyrolitic Carbon). Analisis kuat sumber dan perhitungan awal laju dosis neutron pada teras RDE dilakukan menggunakan program Monte Carlo MCNP5v1.2. Pemodelan heterogenitas ganda pada bahan bakar kernel partikel berlapis TRISO dan pada bahan bakar bola pada teras RDE. Dengan memanfaatkan program EGS99304, jumlah struktur group energi yaitu 640 (SAND-II group structure) digunakan dalam perhitungan spektrum neutron pada reaktor RDE. Teras reaktor RDE dibagi dalam 100 zona (10 arah radial dan 10 arah aksial). Analisis hasil perhitungan menunjukkan bahwa kuat sumber neutron reaktor RDE sebesar 8,47027X1017 neutron/sekon. Distribusi laju dosis neutron ditentukan menggunakan faktor konversi fluks ke dosis neurton dari International Commission on Radiological Protection, ICRP dan NCRP. Hasil perhitungan awal laju dosis neutron dengan faktor konversi ICRP-21 dan NCRP-38 untuk pekerja radiasi pada arah radial di perisai biologis sudah melemah memberikan nilai masing-masing sebesar 6,69915 µSv/jam dan 6,9964 µSv/jam pada posisi 215 cm dari pusat teras RDE, sehingga pekerja radiasi aman dan terlindungi dari radiasi sesuai dengan persyaratan Perka Bapeten  No. 04 tahun 2013 tentang Proteksi dan Keselamatan Radiasi Dalam Pemanfaatan Tenaga Nuklir yang menetapkan nilai batas dosis efektif rerata untuk pekerja radiasi adalah 20 mSv/tahun (10 µSv/jam). Dari hasil analisis tersebut tampak bahwa model perisai radiasi dan perisai biologis telah memenuhi standar keselamatan radiasi yang disyaratkan.Kata kunci: TRISO, Pebble, MCNP5v1.2, RDE, kuat sumber neutron, laju dosis neutron, ICRP, NCRP
CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL Dwijayanto, S.T., R. Andika Putra; Husnayani, Ihda; Zuhair, Zuhair
Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol 25, No 2 (2019): Juni, 2019
Publisher : website

Show Abstract | Download Original | Original Source | Check in Google Scholar | Full PDF (674.731 KB) | DOI: 10.17146/urania.2019.25.2.5525

Abstract

CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There are several options of nuclear fuel utilisation in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilises low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilise other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fuelled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0,627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yield 192.53 curies of radioactivity. These higher radioactivity and photon emission possibly necessitate different measures in spent fuel management, if RDE were to use thorium-based fuel. Tl-208 activity, which found to be emitting potentially non-negligible strong gamma emission, magnified the requirement of proper spent fuel handling especially radiation shielding in spent fuel cask.Keywords: RDE, spent fuel, thorium, HTGR, Tl-208.
Co-Authors Ali Djamhuri Dwijayanto, S.T., R. Andika Putra Hery Adrial, Hery Ihda Husnayani, Ihda Putranto Ilham Yazid Suwoto Suwoto Tagor M. Sembiring, Tagor M.