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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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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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.
Xiang Zhang, Daogui Tian, Yanfang Xue, Lian Chen, Yuquan Li (SNPTC)
Proceedings | Advances in Thermal Hydraulics 2018 | Orlando, FL, November 11-15, 2018 | Pages 1215-1233
In the present study, a two phase transient three-dimensional CFD simulation is carried out to numerically investigate the saturated downward facing pool boiling on a hemispherical heating surface. This simulation is performed within the OpenFOAM framework and a two phase Euler-Euler approach (Volume of Fluid (VOF) model) has been chosen in this research. This code can be used to predict the distribution of the local flow parameters, such as the void fraction, bubble diameter, the velocity of both liquid and gas, the turbulent intensity as well as liquid temperature. Special attention has been devoted to the two phase flow characteristics and vapor morphology along the overheated hemispherical curved surface. For the validation of boiling phenomena, the results of CFD simulation are compared with the visualizations of subscale boundary layer boiling experiment conducted by State Nuclear Power Technology Research & Development Center (SNPTRD) in China. Simulation results show that vapor behavior along the hemispherical curved surface is cyclical, repeatedly forming a stratified vapor layer at the bottom center, which stretches as more vapor is generated, and then flows upwards along from the convex surface. These numerical results show good agreements with the experimental data. Moreover, detailed analysis focusing on the pressure, velocity, void fraction, heating wall temperature as well as the two phase natural circulation flow characteristics are proposed in this paper. It is found that the vapor behavior of downward facing pool boiling on a hemispherical surface is quite different from those on other surface orientations in many respects including the bubble shape, velocity, rise distance and moving trajectory.