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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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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.
M. Chandra Kumar, A. Jasmin Sudha, V. Subramanian, S. Athmalingam, B. Venkatraman
Nuclear Science and Engineering | Volume 197 | Number 1 | January 2023 | Pages 132-143
Technical Paper | doi.org/10.1080/00295639.2022.2103338
Articles are hosted by Taylor and Francis Online.
Melting of the nuclear core is one of the severe accident scenarios in a Sodium-cooled Fast Reactor (SFR). During such an event, molten corium may come into contact with the coolant sodium. This interaction of the molten fuel and the coolant is commonly termed molten fuel–coolant interaction (MFCI) in the nuclear industry. In this study, a numerical analysis is carried out to study the solidification of a molten fuel droplet in the liquid sodium pool. In the first part of the study, the effect of constant internal heat generation on the solidification of the droplet is evaluated with convective heat dissipation prescribed at the droplet surface. The internal heat generation (decay power) and the heat transfer coefficient are varied as parameters, and the time required for complete solidification of the molten droplet is obtained. Based on the results, the freezing of the droplet is categorized into three regimes: conduction limited, transition, and internal heat generation dominated regimes. It is observed that the solidification process of nuclear fuel droplets generated during MFCI is not influenced by internal heat generation and lies in a conduction-limited regime for decay power level prevailing in a medium-sized SFR. Hence, in the next part of the study, the numerical analysis is carried out by incorporating the time-dependent decay power and the temperature-dependent heat transfer coefficient in the computational model by developing user-defined subroutines depicting a realistic scenario of an accident. The results of the analysis show that because of the high subcooling of sodium, film boiling is ruled out; nucleate boiling with a maximum heat transfer rate occurs briefly. The heat transfer coefficient then declines as the interface temperature between the droplet and the sodium decreases rapidly until the natural convective regime is reached. A parametric study on the droplet diameter is also carried out by varying the diameter from 0.5 to 10 mm, spanning the typical particle size spectrum expected during MFCI.
