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Abstract
Validasi perhitungan kekritisan pada Gas Cooled Fast Reactor (GFR) menggunakan kode OpenMC dan SRAC telah dilakukan. OpenMC merupakan kode analisis neutronik yang bersifat open source dan probabilistik yang sedang dikembangkan oleh MIT hingga sekarang. Validasi kode OpenMC perlu dilakukan untuk menunjukkan hasil validitas perhitungan OpenMC dibandingkan dengan kode lainnya. OpenMC yang bersifat probabilistik, mensimulasikan random sampling partikel yang berjumlah besar. Hal terseut bertujuan untuk menunjukkan akurasi perhitungan OpenMC dengan menggunakan partikel yang berjumlah kecil. Validasi dilakukan dengan melihat selisih perhitungan nilai dari kode OpenMC dan SRAC. Nilai konvergensi yang dihasilkan dari kode OpenMC dan SRAC dikatakan tervalidasi dengan memiliki nilai error <1%. Pada penelitian ini menggunakan 50.000 partikel dengan total pengulangan 100 batch aktif dan 30 batch tidak aktif yang disimulasikan. Hasil perbandingan menunjukkan bahwa OpenMC memiliki error maksimal 0,06% terhadap hasil perhitungan kode SRAC
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Copyright (c) 2023 Iklimatul Karomah, Ratna Dewi Syarifah, Nuri Trianti, Artoto Arkundato, Lutfi Rohman, Wenny Maulina, Endhah Purwandari, Umar Sahiful Hidayat
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References
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References
GIF dan DOE Nuclear Energy Research, Analisis Neutronik Pada Reaktor Cepat Dengan Variasi Bahan Bakar (UN-PuN, UC-PuC dan MOX), Jurnal Fisika Unand 3(1), 2002.
GIF dan OECD Nuclear Energy Agency,Technology Roadmap Update for Generation IV Nuclear Energy System, Perancis. Paris, 2004.
Zweifel, P. F, Reactor Physics, USA: McGraw-Hill. Page 1, 2, 10, 74, 1973.
Okumura, K., T. Kugo, K. Kaneko, dan K. Tsuchihashi, SRAC2006: A Comprehensive Neutronics Calculation Code System, Japan: Japan Atomic Energy Agency, 2007.
Mikcus.i,Towards Efficient Monte Carlo Calculations in Reactor Physics,Stockholm,2021.
Syarifah et al, Comparative study on various thermal power for gas cooled fast reactor with uranium plutonium nitride fuel, Cite as: AIP Conference Proceedings, 2020.
Syarifah et al, Comparison of uranium plutonium nitride (U‐Pu‐N) and thorium nitride (Th‐N) fuel for 500 MWth gas‐cooled fast reactor (GFR) long life without refueling, Journal: Energy Reseach, 2017.
Cinantya, D.N dan Fitriyani, D, Analisis Neutronik Pada Reaktor Cepat Dengan Variasi Bahan Bakar (UN-PuN, UC-PuC dan MOX), Jurnal Fisika Unand 3(1), 2014.
Karomah.I, Desain Neutronik Reaktor Cepat Tipe Gfr Berbahan Bakar Uranium Plutonium Karbida Menggunakan Metode One Mesh One Cell. Skripsi Universitas Jember, 2021.
Sabrina et al, Design Study of Gas Cooled Fast Reactor (GFR) with Uranium Plutonium Carbide (UC-PuC) as Fuel with Addition Protactinium (Pa231), Computational and Experimental Research in Materials and Renewable Energy (CERiMRE) Volume 3, Issue 1, page 19-26, 2020.
Ilham, M dan Su’ud,Z, Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel, Journal of Physics: Conference Series, 2017
Ilham et al, Fuel Assembly Design Study for Modular Gas Cooled Fast Reactor using Monte Carlo Parallelization Method, Journal of Physics: Conference Serie, 2018.
Ilham et al,Development of Linkage Program Code OpenMC and ORIGEN2.2 for Neutronic Analysis and Burnup Nuclear Reactor Program. Journal of Physics: Conference Series, 2021.
OpenMC, “The OpenMC Monte Carlo Code,” MIT, 2021. https://docs.openmc.org/en/stable/methods/geometry.html (accessed Jan. 14, 2022).