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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