The use of WIMS-ANL lumped fission product cross sections for burned core analysis with the MCNP Monte Carlo code.
Conference
·
OSTI ID:11044
Most Monte Carlo neutronics analyses are performed for fresh cores. To model snapshots of the cores at different stages during burnup using MCNP, a method is presented that uses lumped fission product (LFP) cross sections generated by the WIMS-ANL code and processed for use in MCNP. Results of analyses for four very different reactor cores using MTR-type and Russian-designed fuel assemblies, with LEU and HEU fuels, are provided to demonstrate the use of this method.
- Research Organization:
- Argonne National Lab., IL (US)
- Sponsoring Organization:
- US Department of Energy (US)
- DOE Contract Number:
- W-31109-ENG-38
- OSTI ID:
- 11044
- Report Number(s):
- ANL/TD/CP-97460; TRN: US0104234
- Resource Relation:
- Conference: 21st International Meeting on Reduced Enrichment for Research and Test Reactors, Sao Paulo (BR), 10/18/1998--10/23/1998; Other Information: PBD: 14 Oct 1998
- Country of Publication:
- United States
- Language:
- English
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