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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
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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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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.
R.-D. Penzhorn, Y. Hatano, M. Matsuyama, Y. Torikai
Fusion Science and Technology | Volume 64 | Number 1 | July 2013 | Pages 45-53
Technical Paper | doi.org/10.13182/FST12-625
Articles are hosted by Taylor and Francis Online.
Stainless steel exposed to gaseous tritium characteristically shows a firmly trapped fraction of tritium in the surface layer, which is not fully removable by water at ambient temperature. Prolonged thermal treatment of tritium-loaded specimens at <443 K causes substantial depletion of the bulk but almost no depletion of the surface layer. For complete removal of hydrogen isotopes from the bulk and the surface, temperatures exceeding 573 K are necessary. Upon chemical etching virtually all tritium trapped in the surface layer appears in the etching solution as tritiated water. Following removal of the layer by chemical etching, the tritium-rich layer reappears after months of aging at ambient temperature with nearly the original tritium activity. Comparison of chronic tritium release rates into liquid water before and after etching reveals that the surface layer only marginally influences the rate. X-ray photoelectron spectroscopy provides evidence that during prolonged aging the surface layer continues to grow while at the same time trapping a fraction of bulk tritium released at ambient temperature. Experimental results suggest different mechanisms of hydrogen uptake and release by the bulk and surface layers. Inference of tritium activity in the bulk of aged or heat-exposed stainless steel material from surface activity measurements may depart significantly from reality.