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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
Meeting Spotlight
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
Standards Program
The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
Latest Magazine Issues
Apr 2025
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Nuclear Science and Engineering
June 2025
Nuclear Technology
Fusion Science and Technology
May 2025
Latest News
Argonne’s METL gears up to test more sodium fast reactor components
Argonne National Laboratory has successfully swapped out an aging cold trap in the sodium test loop called METL (Mechanisms Engineering Test Loop), the Department of Energy announced April 23. The upgrade is the first of its kind in the United States in more than 30 years, according to the DOE, and will help test components and operations for the sodium-cooled fast reactors being developed now.
Jieun Lee, Paolo Balestra, Yassin A. Hassan, Robert Muyshondt, Duy Thien Nguyen, Richard Skifton
Nuclear Technology | Volume 208 | Number 12 | December 2022 | Pages 1769-1805
Technical Paper | doi.org/10.1080/00295450.2022.2081482
Articles are hosted by Taylor and Francis Online.
The verification and validation of Pronghorn is imperative for predicting the fluid velocity, pressure, and temperature in advanced reactors, specifically high-temperature gas-cooled reactors. Pronghorn is a coarse-mesh, intermediate-fidelity, multidimensional thermal-hydraulic code developed by Idaho National Laboratory. The Pronghorn incompressible Navier-Stokes equations are validated by using the pressure drop measurements and axial velocity averaged from the particle image velocimetry data obtained at the engineering-scale pebble bed facility at Texas A&M University.
Pronghorn and STAR-CCM+ porous media models using the Handley, Kerntechnischer Ausschuss, and Carman correlations comparably estimate the pressure drop better than other functions with a maximum 3.34% average relative difference compared to the experimental measurements. The precise average pebble bed porosity estimation has a large impact on the pressure drop. The implementation of the volume-averaged porosity in several sectors, with each sector’s thickness larger than the representative elementary length, has the potential to improve pressure drop modeling or provide more detailed velocity profiles in nuclear reactors with high aspect ratios. The wall effects can be considered using this approach, applying the relatively higher volume-averaged porosity near walls. In addition, the pressure gradients and volume- or surface-averaged axial velocities from the realizable two-layer and shear stress transport models are in good agreement with the porous media simulations and particle image velocimetry data.