Numerical simulations of the turbulence-induced vibrations of a wire-wrapped hexagonal fuel assembly

H. Dolfen, J. de Ridder, L. Brockmeyer, E. Merzari, G. Kennedy, K. van Tichelen, J. Degroote

Research output: Contribution to conferencePaperpeer-review

Abstract

Currently, the Multi-purpose hYbrid Research Reactor for High-tech Applications (MYRRHA) is being designed as a prototype for a fast reactor driven by a particle accelerator. The fuel pins will be separated by wire wrappers and they will be cooled with lead-bismuth eutectic (LBE). Due to this construction and the dense fluid, the occurrence of flow-induced vibrations and possible consequences like fretting have to be investigated. In this research, the focus is on the prediction of the vibration spectrum due to the turbulent flow in the core. The methodology consists of four main steps. First, a modal analysis on a single wire-wrapped fuel pin has been performed to determine its eigenfrequencies and damping ratios in air. Second, a computational fluid-structure interaction (FSI) analysis has been performed on a bundle with seven pins. In these simulations, the wire wrappers have been neglected, but this has been compensated by adjusting the material properties of the pins. These simulations resulted in a spectrum of eigenfrequencies and damping ratios of the fuel pins in contact with LBE. As third step, large eddy simulation (LES) of the flow in the reactor core has been performed, including the wire wrappers and assuming a rigid geometry. This step resulted in time-dependent pressure distributions on the pin surfaces. The final step is then the calculation of the vibration spectrum, by applying these pressure distributions on structural models of the pins with eigenfrequencies and damping ratios corresponding to the average values of those obtained in the second step. The preliminary results predict contact between the fuel pins.

Original languageEnglish (US)
Pages5092-5102
Number of pages11
StatePublished - Jan 1 2019
Event18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019 - Portland, United States
Duration: Aug 18 2019Aug 23 2019

Conference

Conference18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2019
CountryUnited States
CityPortland
Period8/18/198/23/19

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering
  • Instrumentation

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