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Division Spotlight
Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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.
Junbing Zhu, Tianyun Liu, Zhiyuan Ren
Nuclear Science and Engineering | Volume 198 | Number 11 | November 2024 | Pages 2174-2189
Research Article | doi.org/10.1080/00295639.2024.2303171
Articles are hosted by Taylor and Francis Online.
In order to provide a reliable tool for thermal-hydraulic simulation of pebble bed high temperature gas-cooled reactors (HTGRs), a two-dimensional model was developed based on the porous media model and user-defined scalar (UDS) function of FLUENT software. Then, the model was applied to the numerical simulation of the shutdown test of the 10 MW high temperature gas-cooled test reactor (HTR-10) at 9 MW power level, and the temperature distribution and flow field distribution in the reactor were obtained and compared with the results of the experimental data. The reliability of the model in this paper was verified. Based on the model, the effects of the water-cooled panel temperature and the initial core temperature on the thermal-hydraulic characteristics of HTR-10 after shutdown were further explored. The results show that there is a decoupling phenomenon between the residual heat transfer within the core and the heat dissipation of the pressure vessel. The initial core temperature has relatively little effect on the heat dissipation and maximum temperature of the pressure vessel, but it has a significant impact on the thermal characteristics of the core area.