Numerical simulation of supercritical water coolant flow in a GEN IV nuclear reactor by porous media approach
Publish place: Radiation Physics and Engineering، Vol: 2، Issue: 4
Publish Year: 1400
نوع سند: مقاله ژورنالی
زبان: English
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شناسه ملی سند علمی:
JR_RPE-2-4_003
تاریخ نمایه سازی: 9 اسفند 1400
Abstract:
Canadian GEN IV Super Critical Water Reactor (Canadian-SCWR) is a combination version of conventional CANDU reactor with the using super critical water as coolant. Thermal-hydraulic analysis of a nuclear reactor is done to ensure that reactor will work in its safety margins. In this study, thermal hydraulic analysis of Canadian-SCWR is conducted by numerically solving of conservation equations by a porous media approach. The latest concept of Canadian-SCWR core was used for this purpose. In this concept, in each fuel bundles, super critical water flows in two pass and low pressure and low temperature heavy water moderator flows around fuel channel in the Calandria vessel, separately. Average axial temperature, density, heat capacity, pressure and velocity of supercritical water was estimated in two regions of fuel channels (two pass) i.e centeral flow tubes and the fuel rods channel. Compared to the literature, there is a good agreement between our results and the reported results.
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Authors
Danial Salehi
Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran
Gholamreza Jahanfarnia
Department of Nuclear Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran
Ehsan Zarifi
School of Reactor Safety, Nuclear Research and Science Institute, Tehran, Iran
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