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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.
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ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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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 News 40 Under 40 discuss the future of nuclear
Seven members of the inaugural Nuclear News 40 Under 40 came together on March 4 to discuss the current state of nuclear energy and what the future might hold for science, industry, and the public in terms of nuclear development.
To hear more insights from this talented group of young professionals, watch the “40 Under 40 Roundtable: Perspectives from Nuclear’s Rising Stars” on the ANS website.
J. T. Ream, R. P. Varnes
Nuclear Science and Engineering | Volume 13 | Number 4 | August 1962 | Pages 325-337
Technical Paper | doi.org/10.13182/NSE62-A26174
Articles are hosted by Taylor and Francis Online.
It was planned to test full scale U02 test elements in the SRE core. Before doing this, an analysis of the transient behavior of the system in part and the whole was carried out. This analysis concerns the problem of determining transient thermal gradients in the Sodium Reactor Experiment core due to the inability of the after-scram braked flow of the sodium to properly cool the U02 fuel test elements. The analysis showed that the UO2 fuel elements could not be irradiated at the desired core position for maximum power density without exceeding the allowable transient thermal gradient limit. It was necessary to shift them to a position of 25% lower power. An experimental scram of the SRE verified these results for the 19-rod cluster type element. It was possible to concentrate the investigation on the region of the core containing the U02 test elements using the assumption that the steady-state relationship between core pressure drop and reactor flow was valid during flow coastdown. Distributed spatial parameter effects were approximated by a “lumped”-parameter model and were incorporated in sets of coupled finite difference equations which were then solved by use of a general purpose dc analogue computer. The transient flow in the test elements were computed from the SRE quasi-steady-state pressure drop as a function of time. The higher sodium outlet temperature in the U02 test element channels results in an elevation head greater than the elevation head in an SRE channel. This nonlinear buoyant force could not be neglected because it significantly increases the transient flow in the U02 fuel element and stabilizes the channel outlet temperature.