TY - GEN
T1 - Uncoupled DEM Simulation to Investigate the Impact of Coolant flow on Pebble flow in a Pebble Bed Reactor
AU - Malik, Muhammad Sohaib
AU - Chen, Jiaqi
AU - Di Fulvio, Angela
AU - Brooks, Caleb S.
AU - Grunloh, Timothy P.
N1 - Publisher Copyright:
© 2023 Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023. All rights reserved.
PY - 2023
Y1 - 2023
N2 - 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.
AB - 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.
KW - Discrete Element Method (DEM)
KW - Drag
KW - Pebble Bed Reactor (PBR)
KW - Porosity
KW - Residence Time Distribution (RTD)
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U2 - 10.13182/NURETH20-40321
DO - 10.13182/NURETH20-40321
M3 - Conference contribution
AN - SCOPUS:85202977973
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 3238
EP - 3251
BT - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
PB - American Nuclear Society
T2 - 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
Y2 - 20 August 2023 through 25 August 2023
ER -