Experimental determination of flow accelerated corrosion in bend and straight sections of carbon steel primary coolant feeder pipes in pressurised heavy water reactor system

Abstract

Coolant systems in nuclear power plants undergo degradation by various corrosion processes. The carbon steel feeder pipes for the primary heat transport system of pressurised heavy water reactors (PHWRs) undergo extensive wall thinning by a degradation process, known as flow accelerated corrosion (FAC). Flow accelerated corrosion occurs under certain conditions of flow, water chemistry, pipe geometry and material composition. In PHWRs, the outlet feeders carry the heavy water (D₂O) coolant, which flows at velocities in the range of 10 to 20 m s⁻¹. The requirement of high flow velocity at high temperature and pressure to simulate the exact service hydrodynamic conditions, and the long time required to enable unambiguous observations regarding the extent of FAC, makes it difficult to undertake this study in the laboratory. This paper presents the design of an FAC module that enables the relatively rapid measurement of FAC to be undertaken in simulated PHWR primary heat transport conditions. The results of corrosion rate measurements and the attempts to explain the results using a computation fluid dynamics model are presented in this paper.