Abstract
A method of calculating the neutron flux in a heterogeneous system, utilizing a combination of two different energy meshes — coarse and fine — is presented. The method has been programmed into a computer code EXRE-2. This method is particularly suitable for computations of fast reactor spectra that carry both broad and fine undulations. Pointwise cross section curves are prepared by superposition of the Breit-Wigner single level formura. In the energy region where the resonance is not resolved, a random sampling procedure is used for the derivation of resonance parameters. Group averaged resonance capture and fission cross sections and the temperature dependence of these quantities are computed for a typical fast reactor. An examination is made of the applicability of the conventional narrow resonance (NR) approximation applied over a rather wide energy group, and the heterogeneity effect on the resonance cross sections, near the sodium resonance at 2.85 keV. The error due to the NR approximation was found fairly large in the case of the heterogeneous model, but negligible in the homogeneous. The heterogeneity effect was also negligible except for the Doppler cross sections in the immediate neighborhood of 2.85 keV.
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