Lumped Fission Product Neutron Cross Sections Based on ENDF/B-V for Fast Reactor Analysis

The development and evaluation of a lumped fission product neutron cross-section library based on ENDF/B-V data suitable for fast reactor applications have been completed. Both one- and two-lump models have been investigated in detail. Fission product inventories at various burnup steps were calculated by the EPRI-CINDER-2 code and used as weighting functions for lumping. This paper addresses several important issues related to the lumped data including the relative merits of the two models, the dependence on burnup history, the influence of fuel composition and neutron spectrum, the impact of various data bases, the application of the lumped data, the effect of the scattering matrix, and finally the impact on the fission product reactivity worth in a fast reactor system. Although the data and results contained in this paper are specifically related to a particular mixed-oxide core design, they have general validity and application to other similar liquid-metal fast breeder reactor designs.