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Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
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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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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.
Nuclear Technology | Volume 53 | Number 2 | May 1981 | Pages 141-146
Technical Paper | Realistic Estimates of the Consequences of Nuclear Accident / Nuclear Safety | doi.org/10.13182/NT81-A32618
Articles are hosted by Taylor and Francis Online.
A review of the processes important to the behavior of aerosols during a severe reactor accident involving core melting shows processes leading to particle size change (agglomeration, condensation, and evaporation) and processes leading to removal of particles from the atmosphere (diffusion, sedimentation, thermophoretic, and inertial deposition). The NAUA model and computer code developed at the Karlsruhe Nuclear Research Center treats these processes in a hypothetical core melt accident. The NAUA code is based on first principles, without further restrictions. Its application to such an accident in a pressurized water reactor (Biblis B) shows that the mass of aerosol leaked from a containment building during an accident is strongly dependent on the aerosol source from the core and the existing steam conditions. Condensing steam is effective in reducing leaked aerosol mass. Most of the leakage would occur during the first 12 h of an accident; such leakage is not directly proportional to the aerosol source strength but tails off significantly as the initial aerosol concentration increases.