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Division Spotlight
Fusion Energy
This division promotes the development and timely introduction of fusion energy as a sustainable energy source with favorable economic, environmental, and safety attributes. The division cooperates with other organizations on common issues of multidisciplinary fusion science and technology, conducts professional meetings, and disseminates technical information in support of these goals. Members focus on the assessment and resolution of critical developmental issues for practical fusion energy applications.
Meeting Spotlight
Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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
NRC cuts fees by 50 percent for advanced reactor applicants
The Nuclear Regulatory Commission has announced it has amended regulations for the licensing, inspection, special projects, and annual fees it will charge applicants and licensees for fiscal year 2025.
Bastien Faure, Pascal Archier, Jean-François Vidal, Laurent Buiron
Nuclear Science and Engineering | Volume 192 | Number 1 | October 2018 | Pages 40-51
Technical Paper | doi.org/10.1080/00295639.2018.1480190
Articles are hosted by Taylor and Francis Online.
Fast resolution of the Boltzmann transport equation over a nuclear reactor core presupposes the definition of homogenized and energy-collapsed cross sections. In modern sodium fast reactors that rely on heterogeneous core designs, anisotropy in the neutron propagation cannot be neglected, so three-dimensional (3D) models should be used to efficiently compute those effective cross sections. In this paper, the 2D/1D approximation is carried out to overcome computationally expensive 3D calculations while preserving consistent angular representations of the neutron flux. An iterative procedure is defined to solve the 2D/1D equations and produce coarse group homogenized cross sections that account for 3D transport effects. Accuracy of the algorithm is tested on a realistic model of the ASTRID core showing very good results against Monte Carlo simulations for all neutronic parameters (eigenvalue, sodium void worth, and fission map distribution).