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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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.
Masabumi Nishikawa, Kazuya Furuichi, Hiroki Takata
Fusion Science and Technology | Volume 50 | Number 4 | November 2006 | Pages 521-527
Technical Paper | doi.org/10.13182/FST06-A1275
Articles are hosted by Taylor and Francis Online.
Concrete walls play the role not only of the structural material but also of the final barrier of a multiconfinement system of tritium in a fusion reactor or a tritium-handling facility. Therefore, it is required that the behavior of tritium in the concrete materials be clarified to certify the radiation safety of a fusion reactor. The diffusion coefficient of hydrogen in cement paste is obtained by using the permeation experiment in this study, and it is found that the diffusion coefficient of hydrogen in the cement paste is only one order magnitude smaller than the diffusion coefficient of hydrogen in air. Calculation using the diffusion coefficient obtained in this study indicates that the gaseous tritium, HT or T2, can permeate rather rapidly to the outside through the concrete wall of a tritium-handling facility. This calculation implies that installation of a tritium recovery system with proper decontamination performance is required to minimize the tritium transfer to the outer environment.