#TITLE_ALTERNATIVE#
Burnup calculations involving the processing of nuclear data as microscopic cross section. Then macroscopic cross section calculation was done by multiplying the <br /> <br /> <br /> microscopic cross section with a density of nuclides and all materials involved in the reactors i...
محفوظ في:
| المؤلف الرئيسي: | |
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| التنسيق: | Final Project |
| اللغة: | Indonesia |
| الوصول للمادة أونلاين: | https://digilib.itb.ac.id/gdl/view/16317 |
| الوسوم: |
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| الملخص: | Burnup calculations involving the processing of nuclear data as microscopic cross section. Then macroscopic cross section calculation was done by multiplying the <br />
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microscopic cross section with a density of nuclides and all materials involved in the reactors itself. The calculations of diffusion equation was done in two <br />
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dimensional cylindrical coordinates system involving the macroscopic cross section to obtaine the distribution of neutron flux, neutron source density, power density and keff. The calculation gives keff = 1.0017918649. Thus it can be concluded that the reactor is in a supercritical state. After that we use maximum flux to be calculated in the burnup equations to obtain new nuclide density over a 10 years period. In the procces of burnup calculation involved 28 system of differential equation generated from nuclear transmutation chain. It takes a long time and hard to solved analitically. In this thesis the system of differential equation are solved numerically using Fourth Order Runge Kutta Methods to obtain curves of Nuclide Density versus time. |
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