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Three-dimensional nodal diffusion and transport methods for the analysis of fast-reactor critical experiments

Conference ·
OSTI ID:6837556
This paper describes two new nodal methods for solving the multigroup neutron diffusion and transport equations in three-dimensional Cartesian geometry. These methods have been developed for the global analysis of fast-reactor critical experiments once cell-averaged multigroup cross sections for each matrix position or drawer have been computed using appropriate cell-homogenization procedures. Brief descriptions of the nodal diffusion and transport schemes are presented, along with results of two- and three-dimensional calculations for a current Zero Power Plutonium Reactor (ZPPR) configuration.
Research Organization:
Argonne National Lab., IL (USA)
DOE Contract Number:
W-31109-ENG-38
OSTI ID:
6837556
Report Number(s):
CONF-840901-6; ON: DE84011707
Country of Publication:
United States
Language:
English