Analisis Distribusi Fluks Neutron pada Reaktor Berbentuk Slab Menggunakan Persamaan Difusi Multigrup Satu Dimensi dengan Metode Gauss-Seidel

Authors

  • Imra Zakyia Laboratorium Fisika Nuklir, Jurusan Fisika Fakultas Matematika dan Ilmu Pengetahuan Alam, Universitas Andalas
  • Mohammad Ali Shafii Laboratorium Fisika Nuklir, Jurusan Fisika Fakultas Matematika dan Ilmu Pengetahuan Alam, Universitas Andalas

DOI:

https://doi.org/10.25077/jfu.9.3.388-393.2020

Abstract

Telah dilakukan penelitian mengenai distribusi fluks neutron dalam persamaan difusi neutron multigrup satu dimensi.  Jenis reaktor yang digunakan adalah reaktor cepat dengan teras berbentuk slab dan bahan bakar yang digunakan yaitu U-PuN. Penelitian ini menggunakan penampang lintang makroskopik di level sel bahan bakar sebagai masukan awal untuk 70 grup energi.  Data library yang digunakan adalah JFS-3-J33 70 grup energi neutron yang merupakan data dari kode komputer SLAROM dari JAEA Jepang.  Rentang energi dibagi ke dalam tiga daerah grup energi yaitu grup energi cepat, grup energi menengah dan grup energi termal.  Metode iterasi yang digunakan dalam penelitian ini adalah metode iterasi Gauss-Seidel.  Hasil penelitian menunjukkan bahwa distribusi fluks neutron pada grup energi cepat untuk bahan bakar U-235 dan Pu-239 berkisar antara 32,96 n/s cm2  sampai 121,95  n/s cm2, sedangkan pada grup energi menengah terjadi tumpang tindih antar grup energi dan pada grup energi termal distribusi fluks neutron untuk U-238 lebih rendah dibandingkan dengan U-235 dan Pu-239. Perbedaan nilai ini terjadi karena U-238 merupakan bahan fertil. Distribusi fluks neutron pada grup energi cepat memiliki nilai lebih akurat dibandingkan dengan grup energi menengah dan termalkarena penelitian ini didesain untuk reaktor cepat.

 

Research on the distribution of the neutron flux in the one-dimensional multigroup neutron diffusion equation has been done. The type of reactor used is a fast reactor with a slab-shaped reactor core, and the fuel used is U-PuN. The study used macroscopic cross-sections at the fuel cell level as initial input for 70 neutron energy groups. The data library used is JFS-3-J33 70 energy groups, the library data of SLAROM computer codes from JAEA Japan. The energy range is divided into three regions of neutron energy groups, namely fast, medium, and thermal energy groups. The iteration method used in this study is the Gauss-Seidel iteration method. The results showed that the flux distribution in the fast energy group for U-235 and Pu-239 fuels ranged from 32.96 n/s cm2 to 121.95 n/s cm2, whereas in the intermediate neutron energy group overlaps each other and in the thermal energy group the U-238 neutron flux distribution is lower than U-235 and Pu-239. This difference in value occurs because U-238 is fertile material. The distribution of neutron flux in the fast energy group has a more accurate value compared to the medium and thermal energy groups because this study is designed for fast reactors.

References

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Published

2020-08-14

How to Cite

Zakyia, I., & Shafii, M. A. (2020). Analisis Distribusi Fluks Neutron pada Reaktor Berbentuk Slab Menggunakan Persamaan Difusi Multigrup Satu Dimensi dengan Metode Gauss-Seidel. Jurnal Fisika Unand, 9(3), 388–393. https://doi.org/10.25077/jfu.9.3.388-393.2020

Issue

Vol. 9 No. 3 (2020)

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Articles

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