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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.
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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.
J. R. Liaw, H. Henryson II
Nuclear Science and Engineering | Volume 84 | Number 4 | August 1983 | Pages 324-336
Technical Paper | doi.org/10.13182/NSE83-A15453
Articles are hosted by Taylor and Francis Online.
The development and evaluation of a lumped fission product neutron cross-section library based on ENDF/B-V data suitable for fast reactor applications have been completed. Both one- and two-lump models have been investigated in detail. Fission product inventories at various burnup steps were calculated by the EPRI-CINDER-2 code and used as weighting functions for lumping. This paper addresses several important issues related to the lumped data including the relative merits of the two models, the dependence on burnup history, the influence of fuel composition and neutron spectrum, the impact of various data bases, the application of the lumped data, the effect of the scattering matrix, and finally the impact on the fission product reactivity worth in a fast reactor system. Although the data and results contained in this paper are specifically related to a particular mixed-oxide core design, they have general validity and application to other similar liquid-metal fast breeder reactor designs.