An Advanced Neutronic Analysis Toolkit with Inline Monte Carlo capability for BHTR Analysis

Martin, William R. ; Lee, John C.

University of Michigan

关键词: Production;    Geometry;    Fuel Particles;    22 General Studies Of Nuclear Reactors;    Physics;   

DOI  :  10.2172/970985
RP-ID  :  DOE/ID/14745
RP-ID  :  FC07-06ID14745
RP-ID  :  970985

美国|英语

来源: UNT Digital Library

【 摘 要 】

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.