Investigation of 241/243 Am incineration in CANDU reactor using computational method
作者： Z. Gholamzadeh , S. A. H. Feghhi , C. Tenreiro
DOI: 10.1007/S10967-014-3405-6
关键词: MOX fuel 、 Americium 、 Environmental science 、 Incineration 、 Nuclear engineering 、 Spent nuclear fuel 、 Radioactive waste 、 Uranium-233 、 CANDU reactor 、 Thorium
摘要: Americium isotopes are one of the most radiotoxic components nuclear spent fuels. Recently, incineration scenarios for highly elements waste planned by using critical reactors. Computational methods widely used to predict burn up rates such wastes which under fuel matrixes in In this work, MCNPX code has been calculate neutronic behavior an americium-bearing CANDU reactor. Our computational data shows that a thorium-based matrix containing 233U, leads higher americium as well proliferation resistances comparison with other investigated matrix. The contains more negative temperature reactivity coefficients well.
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