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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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Nuclear Science and Engineering
June 2025
Nuclear Technology
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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.
David L. Aumiller, Michael J. Meholic
Nuclear Science and Engineering | Volume 184 | Number 3 | November 2016 | Pages 441-452
Technical Paper | doi.org/10.13182/NSE16-41
Articles are hosted by Taylor and Francis Online.
An assessment of the predictive capability of Coolant Boiling in Rod Arrays–Integrated Environment (COBRA-IE) for critical heat flux (CHF) using the 2005 Groeneveld CHF lookup table is presented. The assessment was performed against 13 different open literature CHF experiments that were conducted over a wide range of conditions in various internal flow geometries. Overall, approximately 1300 data points were evaluated.
Different methodologies to quantify the uncertainty inherent in the CHF models are discussed in this paper. The simulation techniques, uncertainty methods, and results of two of the methods are provided. A discussion of the appropriate use of the CHF uncertainty methods is included. The results indicate that for the method associated with the largest uncertainty, the average measured/predicted value in CHF is 1.19, and the standard deviation is 0.62. For the second method, similar to the critical power ratio used for boiling water reactors, the average ratio is 0.98, and the standard deviation is 0.13. Finally, a method to translate between the methods is proposed and shown to be accurate. The use of this transformation could permit significant time and cost savings by allowing a single uncertainty assessment to serve two very different analytical needs.