Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a computationa...
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| Language: | en |
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2010
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| Online Access: | http://hdl.handle.net/10155/87 |