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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.
Nobuo Sasamoto, Kiyoshi Takeuchi
Nuclear Science and Engineering | Volume 80 | Number 4 | April 1982 | Pages 554-569
Technical Paper | doi.org/10.13182/NSE82-A18969
Articles are hosted by Taylor and Francis Online.
A numerical method is presented for calculating neutron transport problems in three-dimensional (x,y,z) geometry on the basis of a method of direct integration of the integral transport equation. Several new techniques are introduced to the method to make it well adapted to practical neutron transport calculations in three-dimensional geometry. A technique for evaluating the scattering source based on an estimated spectral shape in each material region allows use of coarse energy mesh intervals without reducing calculational accuracy as compared with the calculation with fine meshes. A quadratic function approximation for the source spatial distribution in each spatial mesh interval is found to improve the mathematical error in direct integration of the source term over the spatial variable as compared with the linear- or exponential-function approximation used in the original method. In addition, Lagrange's interpolation formula is applied instead of the linear interpolation used in the original method for more accurate estimation of both flux and source. Comparisons are made of the calculations with experiments for three neutron transport problems, the pool critical assembly experiment, the Winfrith iron benchmark experiment, and the annular duct neutron streaming experiment, and also with the three-dimensional Sn calculation to verify the validity of the present method for neutron transport calculations in (x,y,z) geometry.
