Examination of Multigroup Fission Yield Data UsingPENBURN/PENTRAN Fuel Depletion and Transport Codes

摘要

Compilations of fission yield data incorporated within the PENBURN/PENTRAN depletion code and 3-D Sn parallel transport solver are based on the Los Alamos National Laboratory (LANL) England & Rider fission yield data. PENBURN is a general depletion/burnup code which utilizes zone-based reaction rates obtained through the use of PENTRAN, a parallel multi-group anisotropic Sn code for 3-D Cartesian geometries. A study is performed here to compare nuclide production resulting from the use of multigroup fission data versus the use of single group fission data within PENBURN for a selected set of fission products. Analysis for three modelling scenarios was performed on a Westinghouse PWR fuel unit cell by comparing three group fission yield data against using only thermal group data, and for illustration, using only epithermal group data. Also, similar analysis of multigroup fission data was performed with a fast reactor fuel pin unit cell.