TY - GEN
T1 - Quantifying HFIR Turbulence by Variable Curvature Channels
AU - Mecham, Nicholas J.
AU - Bolotnov, Igor A.
AU - Popov, Emilian L.
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 Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is a very high-flux, pressurized, light water-cooled, flux trap -type research reactor whose current missions are to support neutron scattering experiments, isotope production, materials irradiation, and neutron activation analysis. Because it is difficult to instrument experiments in HFIR's narrow channels, high-fidelity numerical data are needed to verify and calibrate the Reynolds-averaged Navier-Stokes models that are routinely used. Direct numerical simulation (DNS) of the entire spanwise HFIR channel necessitates large, computationally expensive meshes to resolve the turbulence scales because of relatively high velocity flow. It is therefore desirable to determine whether statistically similar turbulence results can be obtained via DNS of smaller domains with comparable curvatures. DNS of coolant flow in several channels of varying curvature was performed at a hydraulic diameter-based Reynolds number of about 70,000. Various locations along the span of the involute HFIR channel were analyzed to assess the effect of curvature on the turbulent flow parameters. Turbulence statistics were collected and compared with HFIR channel results to assess whether a channel with constant curvature and a smaller arc length yields similar results for a location of matching curvature on the HFIR involute geometry. The statistical steadiness of the flow was assessed using conventional total stress relations for flat channels; the applicability of these relations to curved channels is discussed. Mean velocity profiles, turbulent kinetic energy, and turbulence dissipation rates are compared. Higher order turbulence transport terms were calculated and are discussed briefly.
AB - The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is a very high-flux, pressurized, light water-cooled, flux trap -type research reactor whose current missions are to support neutron scattering experiments, isotope production, materials irradiation, and neutron activation analysis. Because it is difficult to instrument experiments in HFIR's narrow channels, high-fidelity numerical data are needed to verify and calibrate the Reynolds-averaged Navier-Stokes models that are routinely used. Direct numerical simulation (DNS) of the entire spanwise HFIR channel necessitates large, computationally expensive meshes to resolve the turbulence scales because of relatively high velocity flow. It is therefore desirable to determine whether statistically similar turbulence results can be obtained via DNS of smaller domains with comparable curvatures. DNS of coolant flow in several channels of varying curvature was performed at a hydraulic diameter-based Reynolds number of about 70,000. Various locations along the span of the involute HFIR channel were analyzed to assess the effect of curvature on the turbulent flow parameters. Turbulence statistics were collected and compared with HFIR channel results to assess whether a channel with constant curvature and a smaller arc length yields similar results for a location of matching curvature on the HFIR involute geometry. The statistical steadiness of the flow was assessed using conventional total stress relations for flat channels; the applicability of these relations to curved channels is discussed. Mean velocity profiles, turbulent kinetic energy, and turbulence dissipation rates are compared. Higher order turbulence transport terms were calculated and are discussed briefly.
KW - Direct Numerical Simulation
KW - HFIR
KW - Turbulent Flow
UR - http://www.scopus.com/inward/record.url?scp=85180635665&partnerID=8YFLogxK
U2 - 10.13182/NURETH20-40044
DO - 10.13182/NURETH20-40044
M3 - Conference contribution
AN - SCOPUS:85180635665
T3 - Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2023
SP - 1194
EP - 1205
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 -