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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.
H. Kamide, K. Hayashi, T. Isozaki, M. Nishimura
Nuclear Technology | Volume 133 | Number 1 | January 2001 | Pages 77-91
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT01-A3160
Articles are hosted by Taylor and Francis Online.
A proper assessment of core thermohydraulics under natural circulation conditions is important so that the full potential of the inherent, passive feature of a fast reactor can be used. When the heat exchangers of the decay heat removal system are operated in the upper plenum of a reactor vessel, cold sodium exiting the heat exchangers may penetrate into the gap regions between fuel subassemblies; this gap flow between the wrapper tubes of subassemblies is called interwrapper flow (IWF). During natural circulation decay heat removal, IWF will significantly modify the flow and temperature distributions in the subassemblies. Sodium experiments were carried out to investigate these phenomena, using a test section consisting of seven subassemblies housed and connected to an upper plenum. The cooling effect of IWF on the fuel subassemblies was evaluated and a new nondimensional parameter was deduced to characterize this effect. On the other hand, IWF reduced the natural circulation flow in the primary loop due to a temperature decrease in the upper part of the core. A balance between the cooling effect and the flow reduction effect is discussed. Three-dimensional analyses were performed to establish an estimation method for IWF. For the temperature decreases due to IWF at the hottest point in the subassemblies there was good agreement between experiments and predictions.