Theoretical and experimental thermal-hydraulic analysis of a new sodium cooled fast breeder reactor core

Abstract

This paper presents the experimental and theoretical results of the thermal-hydraulic design of a new fast breeder reactor core concept. The main feature of this concept is the omission of fuel element cans. The hydraulic function of these fuel element cans is substituted by a winding flow path through the radial blanket and a ring chamber without tubes. A computer code based on the quasi-continuum-theory and especially adapted to the features of the new core concept is developed for theoretical investigations. The pressure drop of the rod bundles is specified by a resistance tensor. The experimental investigations are realized in a test facility, where sodium is simulated by water. Pressures and velocities are measured. Theoretical and experimental results show good agreement. The aim of flattening of the coolant outlet temperature distribution can be reached with satisfying accuracy.