High-fidelity simulation of flow-induced vibrations in helical steam generators for small modular reactors

E. Merzari, H. Yuan, A. Kraus, A. Obabko, P. Fischer, J. Solberg, S. Lee, J. Lai, S. H. Lee, M. Delgado, Y. Hassan

Research output: Contribution to conferencePaperpeer-review

1 Scopus citations

Abstract

Flow-induced vibration is widespread problem in energy systems as they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occur. Given the importance of reliable components in the nuclear industry, flow-induced vibration (FIV) has long been a major concern in the safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from FIV and related failures. In this paper we discuss the use of the computational fluid dynamics code Nek5000 coupled to the structural code Diablo to simulate the flow in helical coil heat exchangers and associated flow induced vibration. In particular, one-way coupled calculations are performed, where pressure and tractions data are loaded into the structural model. The main focus of the paper is on validation of this capability. Fluid-only Nek5000 large eddy simulations are first compared against dedicated high-resolution experiments. Then, one-way coupled calculations are performed with Nek5000 and Diablo for two datasets that provide FIV data for validation. These calculations were aimed at simulating available legacy FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient since the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral pressurized water reactors. We discuss validation of two-way coupled experiments and benchmarks toward the simulation of fluid elastic instability. We briefly discuss the application of these methods to grid-to-rod fretting.

Original languageEnglish (US)
StatePublished - 2017
Event17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017 - Xi'an, Shaanxi, China
Duration: Sep 3 2017Sep 8 2017

Other

Other17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017
CountryChina
CityXi'an, Shaanxi
Period9/3/179/8/17

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering
  • Instrumentation

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