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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.
Dean Wang, Tseelmaa Byambaakhuu
Nuclear Science and Engineering | Volume 195 | Number 12 | December 2021 | Pages 1347-1358
Technical Note | doi.org/10.1080/00295639.2021.1924048
Articles are hosted by Taylor and Francis Online.
It has been well known that the analytic neutron transport solution tends to the analytic solution of a diffusion problem for optically thick systems with small absorption and source. The standard technique for proving the asymptotic diffusion limit is constructing an asymptotic power series of the neutron angular flux in small positive parameter , which is the ratio of a typical mean free path of a particle to a typical dimension of the problem domain. In this paper, first, we provide an analysis of the asymptotic properties of the SN transport eigenvalues. Then, we show that the analytical SN transport solution satisfies the diffusion equation in the asymptotic diffusion limit based on a recently obtained closed-form analytical solution to the one-dimensional monoenergetic SN neutron transport equation. The boundary conditions for the diffusion equation are discussed.