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Mathematics & Computation
Division members promote the advancement of mathematical and computational methods for solving problems arising in all disciplines encompassed by the Society. They place particular emphasis on numerical techniques for efficient computer applications to aid in the dissemination, integration, and proper use of computer codes, including preparation of computational benchmark and development of standards for computing practices, and to encourage the development on new computer codes and broaden their use.
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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.
C. R. Gould, A. I. Hawari, E. I. Sharapov
Nuclear Science and Engineering | Volume 165 | Number 2 | June 2010 | Pages 200-209
Technical Paper | doi.org/10.13182/NSE09-48
Articles are hosted by Taylor and Francis Online.
We revisit the determination by Bowman et al. of unusual neutron transport characteristics for a newly fabricated form of graphite [Nucl. Sci. Eng., 159, 182 (2008); Nucl. Sci. Eng., 161, 68 (2009)]. From MCNP modeling and consideration of data from other experiments, we determine revised values for the neutron transport parameters of this graphite. Our reanalysis gives a coherent scattering cross section coh ˜ 4 b at 50 meV, a small-angle neutron scattering cross section sans ˜ 11 to 13 b at 1 meV, and an effective capture cross section a = 5.8 ± 0.5 mb. Scaled to a graphite reference density of 1.60 g/cm3 , we find a diffusion coefficient [overbar D] = 0.94 ± 0.03 cm and a diffusion length L = 47.7 ± 3.7 cm. Apart from the somewhat larger values of a and [overbar D], these are not untypical parameters for graphite. Based on our investigation, the recent experiments and analysis of Bowman et al. do not give evidence for different transport properties for this newly fabricated graphite.
