ЩҶЩҲШ№ Щ…ЩӮШ§Щ„ЩҮ : Щ…ЩӮШ§Щ„ЩҮ ЩҫЪҳЩҲЩҮШҙЫҢ

ЩҶЩҲЫҢШіЩҶШҜЪҜШ§ЩҶ

ЪҜШұЩҲЩҮ ЩҒЫҢШІЫҢЪ©ШҢ ШҜШ§ЩҶШҙЪ©ШҜЩҮ Ш№Щ„ЩҲЩ…ШҢ ШҜШ§ЩҶШҙЪҜШ§ЩҮ Ш§ШұЩҲЩ…ЫҢЩҮШҢ ШөЩҶШҜЩҲЩӮ ЩҫШіШӘЫҢ: 165-57153ШҢ Ш§ШұЩҲЩ…ЫҢЩҮ ЩҖ Ш§ЫҢШұШ§ЩҶ

Ш№ЩҶЩҲШ§ЩҶ Щ…ЩӮШ§Щ„ЩҮ [English]

Simulation of Neutron Flux Distribution in a Cylindrical Critical Heterogeneous Reactor with Different Fuel Concentrations, Using Finite Element Method (FEM)

ЩҶЩҲЫҢШіЩҶШҜЪҜШ§ЩҶ [English]

• R Khoda-Bakhsh
• S Behnia
• A Jafari

The finite element method is applied to the spatial variables of multi-group neutron transport equation in a two-dimensional cylindrical (r, z) geometry. The equation is discretized using rectangular sub regions in the (r, z) plane. The discontinuous method with the bilinear or biquadratic Lagrang's interpolating polynomials and basis functions is used in the ANSYS program. Here, the angular fluxes are allowed to be discontinued across the sub region boundaries. Some numerical calculations have been made on a real cylindrical Aristotle reactor with different fuel concentrations on the fuel rods; the results indicate that the flux and power of the heterogeneous critical reactor increase on the edges of the core in comparison with the homogeneous one.

• Finite element method
• Fuel Concentration
• Cylindrical Reactor
• Heterogeneous Reactor
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