Application of TULIP/STREAM code in 2-D Fast Reactor Core Neutronic Analysis

Abstract

A deterministic MOC code STREAM of the Computational Reactor Physics and Experiment (CORE) laboratory of Ulsan National Institute of Science and Technology (UNIST), was initially designed for the calculation of pressurized water reactor two- and three- dimensional assemblies and cores. Since fast reactors play an important role in the generation-IV concept, it was decided to upgrade the code for the analysis of fast neutron spectrum reactors. This paper presents a coupled code TULIP/STREAM developed for the fast reactor assembly and core calculations. The TULIP code produces self-shielded multi-group cross-sections using a one-dimensional cylindrical model. The generated cross-section library is used in the STREAM code which solves eigenvalue problems for a twodimensional assembly and a multi-assembly whole reactor core. Multiplication factors and steady-state power distributions were compared with the reference solutions obtained by a continuous energy Monte-Carlo code MCS. With the developed code, sensitivity study of the number of energy groups, the order of anisotropic Pn scattering, and the multi-group cross-section generation model was performed on the keff and power distribution. The 2D core simulation calculations show that the TULIP/STREAM code gives k-eff error smaller than 200 pcm and the root mean square errors of the assembly and pin-wise power distributions about 0.13% and 0.27%, respectively.