1. Calculation of neutron flux distribution and diffusion coefficient in the slab reactor core using multi-group diffusion equation with Jacobi method.
- Author
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Shafii, Mohammad Ali, Muldarisnur, Tongkukut, Seni Herlina Juita, and Arkundato, Artoto
- Subjects
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NEUTRON flux , *JACOBI method , *HEAT equation , *DIFFUSION coefficients , *NUCLEAR reactor cores , *NEUTRON sources - Abstract
In general, the neutron transport equation cannot be solved analytically. One of the ways to solve the neutron transport equation is to use the multi-group diffusion equation. This study uses the Jacobi method to determine the neutron flux distribution and the neutron diffusion coefficient using the multi-group diffusion equation in a slab reactor core. The type of reactor used in this study is a fast reactor with Uranium-Plutonium Nitride (U-PuN) nuclear fuel. The data library used in this study is the SLAROM data library, namely JFS-3-J33 from JAEA (Japan Atomic Energy Agency). Flux distribution and neutron diffusion coefficient were calculated for 70 neutron energy groups by dividing the fast and the thermal energy group. The results showed that the distribution of neutron flux due to neutron interactions in the U-235 and Pu-239 fissile materials in the fast energy group region was more significant than in the thermal energy group region. Different results were obtained for the fertile material U-238, where the value of the neutron flux distribution was the same in the fast and thermal energy groups. The neutron diffusion coefficient for nuclear fuel U-235 and Pu-239 are low in the high energy group and high in the thermal energy group. This phenomenon occurs because nuclear fuels of U-235 and Pu-239 are fissile materials. However, this does not apply to U-238 nuclear fuel because U-238 is a fertile material. U-238 experiences the lowest neutron diffusion coefficient in the fast and thermal energy groups. [ABSTRACT FROM AUTHOR]
- Published
- 2024
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