Uncoupled DEM Simulation to Investigate the Impact of Coolant flow on Pebble flow in a Pebble Bed Reactor

Muhammad Sohaib Malik, Jiaqi Chen, Angela Di Fulvio, Caleb S. Brooks, Timothy P. Grunloh

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

Abstract

The pebble bed reactor (PBR) is a type of high-temperature gas-cooled reactor (HTGR) among the generation-IV nuclear reactor designs.TRISO-fueled pebbles flow through the reactor and leave at a controlled discharge rate.High-pressure helium gas flows from top to bottom through gaps between pebbles as the coolant.Understanding pebble and fluid flow interaction are essential to accurately predict the thermal-hydraulic characteristics of a PBR.Previously, pebble flow characteristics such as residence time distribution have been measured using radioactive particle tracking (RPT).However, there is a lack of experimental data and numerical results showing the impact of the coolant flow on the pebble flow.Local characterization of pebble flow helps calculate the fluid drag force and understand its impact on pebble flow.The discrete Element Method (DEM) has found applications in understanding granular particle motion.This work uses the LIGGGHTS software to run uncoupled particle flow simulations under gravity in recirculation mode.The drag force on each particle is explicitly calculated based on the Schiller-Naumann correlation.Global and local pebble parameters such as Residence Time Distribution (RTD), the spatial distribution of vertical velocity, porosity, and volume fraction, are inspected for two cases with and without drag force, to check the drag force impact on particles.The drag force model used showed no considerable difference between the pebble flow characteristics for the two cases.An improved implementation of the drag force model based on correlations that incorporate local variation in flow characteristics could help in a better understanding of the effect of fluid-pebble interaction force on pebble flow.

Original languageEnglish (US)
Title of host publicationProceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PublisherAmerican Nuclear Society
Pages3238-3251
Number of pages14
ISBN (Electronic)9780894487934
DOIs
StatePublished - 2023
Externally publishedYes
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

Keywords

  • Discrete Element Method (DEM)
  • Drag
  • Pebble Bed Reactor (PBR)
  • Porosity
  • Residence Time Distribution (RTD)

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Instrumentation

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