High-Fidelity CFD Simulation of Mixed Convection in a Pebble Bed Test Reactor Core

Haomin Yuan, Dillon Shaver, Tri Nguyen, Elia Merzari, David Reger, Ka Yen Yau, Giacomo Busco, Nate Salpeter, R. Brian Jackson

Research output: Chapter in Book/Report/Conference proceedingConference contribution

1 Scopus citations

Abstract

The Hermes low-power (35 MWth) reactor will be built and operated by Kairos Power LLC to demonstrate their Fluoride cooled high-temperature reactor (KP-FHR) technology. In the KP-FHR, the reactor core is composed of randomly packed pebbles with TRISO fuel particles inside. FLiBe flows upward through the pebble bed acting as a coolant to remove heat. Under low to moderate flow conditions, buoyancy plays a critical role. In this study, we simulated a pebble bed core with 34374 pebbles randomly packed, similar to the Hermes reactor's size. The core radius is 14 times of the pebble-diameter, while the core height is 45 times. A Large-Eddy Simulation (LES) of the flow inside the pebble bed has been performed at Re = 160 (based on inlet velocity and pebble diameter) and pebble power at 1680 W per pebble. The Spectral Element CFD code NekRS with GPU acceleration was used for this study. The low-Mach number approximation has been applied to address property changes in the FLiBe and account for buoyancy. A pure hexahedral mesh with 60 million elements was generated by the Voronoi-cell method. The Grashof number in the simulation is on the order from 106 to 107. The local Reynolds number is on the order of 103, resulting in a Richardson number between 1 and 10 across the domain. Thus, the flow is in the mixed convection regime. The Nusselt number across the domain is obtained and compared with empirical correlations.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages645-658
Number of pages14
ISBN (Electronic)9780894487934
DOIs
StatePublished - 2023
Event20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023 - Washington, United States
Duration: Aug 20 2023Aug 25 2023

Publication series

NameProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023

Conference

Conference20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Country/TerritoryUnited States
CityWashington
Period8/20/238/25/23

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering
  • Instrumentation

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