Flow-induced Vibration Analysis of Steam Generators and Fuel Assemblies With the VIBIC Computer Code

Abstract

Abstract Flow-induced vibration and consequent fretting wear have always been important factors to be considered in the design and operation of nuclear reactors. Due to significant changes in mechanical design and thermalhydraulic conditions, these topics are at least as important for advanced nuclear reactors, including future small modular reactors. Atomic Energy of Canada Limited (AECL) has long been conducting studies to understand and solve flow-induced vibration and fretting-wear problems in nuclear reactors. As part of this work, AECL developed the computer codes VIBIC and PIPO-FE. In use since the 1970’s, the codes have been continually updated and now are used primarily in the operational assessment and design of fuel assemblies and steam generators. Recently, the VIBIC code was used to assess the expected vibration and fretting wear performance of a Gen-IV fuel assembly to be used in a fuel qualification test. The assessment showed that VIBIC is capable to provide initial wear predictions of damage due to flow induced vibration in Gen-IV fuel. The predictions will be refined once the fluid forcing function has been verified and material wear properties have been measured, at expected Gen IV conditions. This paper reviews AECL’s existing flow-induced vibration assessment technology and identifies how the fluid forcing function and study of material wear properties need to be extended to address fuel vibration in supercritical water reactors. Keywords: Safety, vibration analysis, fretting-wear analysis, computer code, flow-induced vibration, supercritical reactor, nuclear fuel.