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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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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.
Erik Bodmann, Hans-JüRgen Breuer, Gerhard Raule, Manfred Rödig
Nuclear Technology | Volume 66 | Number 3 | September 1984 | Pages 667-674
H. Design Codes and Life Prediction | Status of Metallic Materials Development for Application in Advanced High-Temperature Gas-Cooled Reactor / Material | doi.org/10.13182/NT84-A33488
Articles are hosted by Taylor and Francis Online.
Studies of material behavior under complex loading form a bridge between standard material testing methods and the stress analysis calculations for reactor components at high temperatures. The aim of these studies is to determine the influence of typical load change sequences on material properties, to derive the equations required for stress analyses, to carry out tests under multiaxial conditions, and to investigate the structural deformation mechanisms of creep buckling and ratcheting. The present state of the investigations within the high-temperature gas-cooled reactor materials program is described, with emphasis on the experimental apparatus, the scope of the program, and the initial results obtained.