An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis
- University of Michigan/Department of Nuclear Engineering
Monte Carlo capability has been combined with a production LWR lattice physics code to allow analysis of high temperature gas reactor configurations, accounting for the double heterogeneity due to the TRISO fuel. The Monte Carlo code MCNP5 has been used in conjunction with CPM3, which was the testbench lattice physics code for this project. MCNP5 is used to perform two calculations for the geometry of interest, one with homogenized fuel compacts and the other with heterogeneous fuel compacts, where the TRISO fuel kernels are resolved by MCNP5.
- Research Organization:
- University of Michigan
- Sponsoring Organization:
- USDOE
- DOE Contract Number:
- FC07-06ID14745
- OSTI ID:
- 970985
- Report Number(s):
- DOE/ID/14745
- Country of Publication:
- United States
- Language:
- English
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