Neutronic Analysis For Nuclear Reactor Systems pp 213-252 | Cite as
Energy Effects in Modeling Neutron Diffusion: Two-Group Models
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Abstract
In this chapter, we derive the multi-group diffusion equation (MGDE) and we illustrate how do we solve them in a way that allows us to calculate an accurate eigenvalue and accurate reaction rates. Since the cross sections vary wildly by multiple orders of magnitude over the energy range in a typical nuclear reactor, the major problem becomes one of determining accurate multi-group cross sections for the design problem under consideration.
References
- 1.J.J. Duderstadt, L.J. Hamilton, Nuclear Reactor Analysis (Wiley). 1976 editionGoogle Scholar
- 2.P.F. Zweifel, Reactor Physics (McGraw-Hill, New York, 1973)CrossRefGoogle Scholar
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