ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
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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!
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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.
Seungwon Shin, S. I. Abdel-Khalik
Nuclear Science and Engineering | Volume 156 | Number 1 | May 2007 | Pages 24-39
Technical Paper | doi.org/10.13182/NSE07-A2682
Articles are hosted by Taylor and Francis Online.
The behavior of an evaporating thin liquid film on a nonuniformly heated cylindrical rod with both parallel and cross vapor flow has been numerically investigated. The aim is to develop a mechanistic model for local dryout in boiling water reactors (BWRs). The liquid film on a full-length BWR fuel rod may experience significant axial and azimuthal heat flux gradients and cross flow due to variations in the thermal-hydraulic conditions in surrounding subchannels caused by proximity to an inserted control blade tip and/or the top of part-length fuel rods. Such heat flux gradients coupled with localized cross flow may cause the liquid film on the fuel rod surface to rupture by hydrodynamic instability, thereby forming a dry hot spot. These localized dryout phenomena cannot be accurately predicted by traditional subchannel analysis methods in conjunction with empirical dryout correlations. To this end, a numerical model based on the level contour reconstruction method has been developed. The model includes a ghost-cell extrapolation technique to handle the complex interface geometry. Additionally, a sharp interface temperature technique has been implemented. Application of the model to BWR fuel rods shows that localized cross flow coupled with heat flux gradients can lead to liquid film rupture and dry spot formation.