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- W2488142734 abstract "The integro-differential equation describing the spatially dependent neutron balance within a nuclear reactor, while formally correct, is extremely challenging to solve in most situations. Fortunately, this equation may be considerably simplified by using the diffusion theory approximation discussed earlier and by averaging the energy-dependent neutron cross sections over discrete energy regimes (or energy groups). The set of equations which result from these simplifications are called the multigroup neutron diffusion equations. These equations consist of coupled ordinary differential equations with constant coefficients which may be solved relatively easily using standard mathematical techniques. The solution to these equations yields spatially dependent neutron flux distributions for each discrete energy group plus an eigenvalue which describes the critical state of the reactor. By using these neutron flux distributions with the appropriate fission cross sections, detailed, fairly accurate power density distributions may be determined at all points throughout a nuclear reactor." @default.
- W2488142734 created "2016-08-23" @default.
- W2488142734 creator A5062613393 @default.
- W2488142734 date "2016-01-01" @default.
- W2488142734 modified "2023-09-23" @default.
- W2488142734 title "Multigroup Neutron Diffusion Equations" @default.
- W2488142734 doi "https://doi.org/10.1016/b978-0-12-804474-2.00008-4" @default.
- W2488142734 hasPublicationYear "2016" @default.
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