ANALISIS NEUTRONIK KEKRITISAN TERAS REAKTOR NUSCALE BERBAHAN BAKAR DENGAN MENGGUNAKAN SOFTWARE OPENMC
DOI:
https://doi.org/10.22437/jop.v8i1.20812Keywords:
Keywords: Burn-up, effective multiplication factor, fission reaction rate, Neutron flux distribution, OpenMC.Abstract
[Title: Neutronic Analysis of the criticality level of the nuscale reactor core with fuel using the openMC software] This study analyzed the design of the NuScale reactor, which aims to determine the level of criticality by modeling the shape of the cell pin, assembly, and core with nuclear fuel in the form of uranium dioxide, which will be varied by changing the percentage of uranium-235 content as much as 0% to 7% by using monte carlo methods in OpenMC program code. This study was conducted to obtain the design of nuclear reactors as well as the calculation of the effective multiplication factor, fission reaction rate, and neutron flux distribution for two years of the combustion process (Burn-up). The result of the calculation for the effective multiplication factor and reaction rate states that the greater the percentage of enrichment in uranium-235, the greater the value of the resulting in both parameters. While the distribution of neutron flux produces the most significant value in the middle region or center of the fuel and is seen from the average value produced, and the smallest value is at the edge of the cell. The analysis of this NuScale reactor can later be used as a reference in preparing a safe and efficient reactor core.
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