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Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR.
(2014-08-01)
The neutron flux throughout a reactor core should be calculated, ideally, by solving the neutron transport equation for a highly detailed geometric model of the core. Since this is computationally impractical, approximate ...
Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices
(2010-04-01)
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 ...