Argonne National Laboratory
Reduced Enrichment for Research and Test Reactors
Nuclear Science and Engineering Division at Argonne
U.S. Department of Energy

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Abstracts and Available Papers Presented at the
1998 International RERTR Meeting

The Use of WIMS-ANL Lumped Fission Product Cross Sections for
Burned Core Analysis with the MCNP Monte Carlo Code

N. A. Hanan, R. B. Pond, W. L. Woodruff, M. M. Bretscher, and J. E. Matos
Argonne National Laboratory
Argonne, IL, USA


Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the core at different stages during burnup using MCNP, a method is presented that uses lumped fission product cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for several research reactor cores using MTR and Russian-designed fuel assemblies are provided to demonstrate the use of this method.

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Dr. Nelson Hanan
Nuclear Engineer
Argonne National Laboratory
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Last modified on July 29, 2008 11:33 +0200