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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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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. P. Sharma, A. K. Nayak
Nuclear Technology | Volume 197 | Number 2 | February 2017 | Pages 158-170
Technical Paper | doi.org/10.13182/NT15-127
Articles are hosted by Taylor and Francis Online.
The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube–type, heavy water–moderated, and boiling light water–cooled natural-circulation–based reactor. The fuel bundle of AHWR contains 54 fuel rods arranged in three concentric rings of 12, 18, and 24 fuel rods. This fuel bundle is divided into a number of imaginary interacting flow passages called subchannels. Transition from a single-phase-flow condition to a two-phase-flow condition occurs in the reactor rod bundle with increase in power. Prediction of the thermal margin of the reactor has necessitated the determination of intersubchannel mixing due to void drift. Void drift is due to redistribution of the non-equilibrium void fraction to attain an equilibrium void fraction. This redistribution occurs in the reactor rod bundle until it reaches the state of equilibrium void fraction. Hence, it is vital to evaluate void drift between subchannels of AHWR rod bundles.
In this paper, experiments were carried out to investigate the void drift phenomena in simulated subchannels of AHWR. The size of the rod and the pitch in the test section were the same as those of the actual rod bundle in the prototype. Three subchannels are considered in 1/12th of the cross section of the rod bundle. Water and air were used as the working fluid, and the experiments were carried out at atmospheric condition without the addition of heat. The void fraction in the simulated subchannels was varied from 0 to 0.8 under various ranges of superficial liquid velocities. The void drift between the subchannels was measured. The test data were compared with existing models in the literature. It was found that the existing models could predict the measured equilibrium void fraction in the rod bundle of the reactor within the range +8% to −14%.