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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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
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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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.
Joel L. McDuffee, Arthur E. Ruggles
Nuclear Science and Engineering | Volume 125 | Number 2 | February 1997 | Pages 232-242
Technical Paper | doi.org/10.13182/NSE97-A24270
Articles are hosted by Taylor and Francis Online.
A model is presented for predicting the pressure gradient in partially developed subcooled boiling of water for velocities from 15 to 30 m/s and inlet peaked, nonuniform axial flux profiles with channel average flux values of 6 MW/m2. The partially and fully developed boiling regions are considered separately, however; the same general modeling technique is used for both. Several correlations for the void fraction at onset of significant void are considered, and their effect on the channel pressure drop is evaluated. The effect of nonuniform axial heat flux on the channel pressure drop is also evaluated. The model is compared with pressure drop data from the thermal-hydraulic test loop at Oak Ridge National Laboratory and is found to agree with the data within 24%.
