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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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Fusion Science and Technology
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.
Eric Lang, Chase N. Taylor, Nathan Madden, Trevor Marchhart, Charles Smith, Xing Wang, Jessica Krogstad, J. P. Allain
Fusion Science and Technology | Volume 79 | Number 5 | July 2023 | Pages 592-601
Technical Paper | doi.org/10.1080/15361055.2022.2164444
Articles are hosted by Taylor and Francis Online.
Tungsten is the material of choice for plasma-facing components in the divertor region of future nuclear fusion reactors. Exposure to low-energy helium ion irradiation results in microstructural changes as helium is trapped at defects in the tungsten matrix. High-temperature exposure results in the formation of helium bubbles in the subsurface. Dispersion-strengthened tungsten materials are tungsten-based materials with added transition metal carbides to alter the impurity distribution and grain structure. In this work, the thermal release of helium from dispersion-strengthened tungsten is investigated. After irradiation at 1073 K to a 1024 m−2 fluence, thermal desorption spectroscopy was performed to elucidate the helium trapping and desorption behavior. Post-desorption microscopy was performed to correlate the microstructural changes with helium release spectra. The amount of desorbed helium was highest in the 1.1 and 5 wt% alloys, and significantly lower in the 10 wt% alloys. Helium bubbles were observed in the pure tungsten and 1.1 wt% alloys within the tungsten grains. Correlating the composition with helium release spectra revealed the importance of tailoring grain size and oxide vacancy concentrations by varying the dispersoid content on the helium retention and release behavior. These first results of helium desorption from dispersion-strengthened tungsten indicate compositionally dependent retention and reveal the need to examine helium retention in advanced tungsten alloys under reactor-relevant exposure.