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Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
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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
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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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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.
Efigenio Cubillos-Moreno, Mohamed Belhadj, Tunc Aldemir
Nuclear Technology | Volume 98 | Number 3 | June 1992 | Pages 333-348
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT92-A34663
Articles are hosted by Taylor and Francis Online.
The heat flux that leads to onset of nucleate boiling qONB is an important quantity for plate-type research reactors since it is frequently used as a thermal design constraint and also indicates the transition point from single- to two-phase heat removal in transient analyses. Recent experimental work has shown that qONB can be sensitive to both channel gap size d and flow velocity v under laminar, upward flow conditions that are encountered in such reactors under naturalconvection core cooling. New experimental data are presented to test the validity of the correlation proposed from the results of the previous work in extended d and local pressure p ranges. The correlation predicts the new experimental data for mixed or pure buoyancy-driven upward flows in 2.0 ≤ d ≤ 5.0 mm channels with 0.03 ≤ v ≤ 0.16 m/s and 1.05 × 105 ≤ p ≤ 1.70 × 105 Pa within 25%. The new d range covers almost all the existing and planned plate-type research reactors. The p range extends the applicability of the correlation to the analysis of a number of accident scenarios in open-pool reactors with power levels up to 5 to 10 MW, such as partial loss of pool water or coolant pump trip. The pressure range is also relevant to the analysis of similar accidents in higher power pressurized systems if the accident is accompanied by system depressurization. In the implementation of the correlation for such analyses, it is important to note that the correlation implicitly assumes that the wall superheat is nonnegative.
