Serpent Monte-Carlo Code: An Advanced Tool for Few-Group Cross Section Generation

Serpent is a continuous-energy Monte Carlo (MC) reactor physics code recently developed at VTT Technical Research Centre of Finland(1). Serpent can be used for 2D fuel lattice calculations as well as for 3D full core simulations.Serpent is especially designed to generate homogenized constants for deterministic 3D core analysis. For any region of interest the code automatically calculates homogenized few-group cross sections, group-to-group scattering matrices, diffusion coefficients, assembly discontinuity factors, kinetics parameters, etc.
The capability to generate homogenized few-group constants can be considered as one of the most attractive features of Serpent. Being a MC code, Serpent is capable of handling complex geometries without any major approximations and can be used for producing cross section data for virtually any fuel or reactor type. The demonstration of the Serpent capability to generate few-group cross sections for different reactor systems is the main topic of this paper.