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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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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.
C. Fagan, M. Sharpe, W. T. Shmayda, W. U. Schröder
Fusion Science and Technology | Volume 75 | Number 8 | November 2019 | Pages 1058-1063
Technical Paper | doi.org/10.1080/15361055.2019.1610308
Articles are hosted by Taylor and Francis Online.
In this work, Aluminum 6061-T6 samples were subjected to MIL-DTL-5541F type-I, class-3 anodic coatings, where a yellow irradiate finish was achieved. Both chromate-conversion coatings (CCCs) and unmodified samples were exposed to deuterium-tritium (PT = 0.51 atm) gas for 24 h at room temperature. Following loading, the samples were subjected to one of two desorption techniques: temperature-programmed desorption or a surface stripping technique. The results show that chromic-acid anodizing of aluminum dramatically increases the total quantity of tritium retained by the treated surface as compared to unmodified aluminum. X-ray photoelectron spectroscopy and scanning electron microscopy studies of both treated aluminum and unmodified samples indicate that the CCCs contain significant quantities of hydrated chromium. Using transmission electron microscopy, the surface is shown to have significant cracking and fracturing of the film and leads to a highly grained and porous surface. Such surface defects coupled with the vast quantity of hydration sites are likely reasons for the increased retained tritium inventory observed for CCC samples. Because of the physical and chemical properties of unmodified CCC samples, they are not suitable for use in tritium environments.