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Division Spotlight
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.
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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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.
Nnaemeka Nnamani
Nuclear Science and Engineering | Volume 198 | Number 10 | October 2024 | Pages 1950-1957
Research Article | doi.org/10.1080/00295639.2023.2284453
Articles are hosted by Taylor and Francis Online.
The results of the thermalized flux calculation that incorporate radiative capture reactions in the presence and absence of polyethylene blocks that form an enclosure for a deuteron-deuteron (D-D) neutron generator are presented. This method can be used to measure the moderated neutron flux component in a mixture of moderated and primary neutron spectra. Using 20-cm-thick polyethylene blocks to surround a D-D neutron generator, the moderation of primary neutrons was investigated using nine indium foils. In this paper, the relationship between the moderated neutron flux and the radiative capture rates in the presence and absence of polyethylene blocks is derived. This is compared to a MCNP simulation and a calculation of modulated flux that ignore the primary neutron components.
