Core thermal hydraulic CFD support for liquid metal reactors

F. Roelofs*, D. Dovizio, H. Uitslag-Doolaard, D. De Santis, A. Mathur, B. Mikuz, A. Shams

*Corresponding author for this work

Research output: Contribution to journalArticlepeer-review

11 Scopus citations

Abstract

Knowledge of the heat transport in the core is important for design and safety assessment of all nuclear reactors including liquid metal cooled reactors. In the past, design and safety calculations with respect heat transport in the core for such liquid metal cooled reactors were largely one-dimensional and based on experimental data. Nowadays, with modern state-of-the-art computer power and tools, three-dimensional Computational Fluid Dynamics (CFD) simulations allow designers and safety specialists to obtain much more detailed information on the heat transport in liquid metal cooled fuel assemblies, obviously supported by necessary experimental campaigns. This may lead to new insights possibly decreasing the safety margins. To this respect, an overview will be provided on the necessary activities in the frame of design and safety support using CFD for liquid metal reactors accompanied and illustrated by examples from NRG in the Netherlands. These examples include validation efforts for fuel assemblies as they are designed on the drawing board for ‘cold’ conditions. However, in reality, even under normal operational condition, a fuel assembly may deform. Therefore, an assessment of the effect of deformations resulting from operational conditions is necessary and will be shown. Another aspect possibly occurring during operational conditions is vibration. State-of-the-art coupled CFD and finite element method fluid structure interaction techniques have been developed and applied to a wire wrapped fuel assembly, providing insights in the vibration behavior of such assemblies. However, design and safety analysts will not only have to cope with operational conditions, but also have to show the heat transport behavior under accident conditions. For this, an assessment of the effect of internal and inlet blockages will be presented.

Original languageEnglish
Article number110322
JournalNuclear Engineering and Design
Volume355
DOIs
StatePublished - 15 Dec 2019
Externally publishedYes

Bibliographical note

Publisher Copyright:
© 2019 Elsevier B.V.

Keywords

  • CFD
  • Fuel assembly
  • Single-phase
  • Wire-wraps

ASJC Scopus subject areas

  • Nuclear and High Energy Physics
  • Nuclear Energy and Engineering
  • General Materials Science
  • Safety, Risk, Reliability and Quality
  • Waste Management and Disposal
  • Mechanical Engineering

Fingerprint

Dive into the research topics of 'Core thermal hydraulic CFD support for liquid metal reactors'. Together they form a unique fingerprint.

Cite this