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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
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.
Quentin Newell, Charlotta Sanders
Nuclear Science and Engineering | Volume 179 | Number 3 | March 2015 | Pages 253-263
Technical Paper | doi.org/10.13182/NSE13-44
Articles are hosted by Taylor and Francis Online.
The Monte Carlo (MC) method is becoming popular for three-dimensional fuel depletion analyses to compute quantities of interest in used nuclear fuel including isotopic compositions. However, there are some questions concerning the effect of MC uncertainties on predicted results in MC depletion calculations. The MC method introduces stochastic uncertainty in the computed fluxes. These fluxes are used to collapse cross sections, estimate power distributions, and deplete the fuel within depletion calculations; therefore, the predicted number densities also contain random and propagated uncertainties due to the MC solution to the neutron transport equation. The linear uncertainty nuclide group approximation (LUNGA) method was developed to calculate the propagated stochastic uncertainty in the nuclear isotopics, using the time-varying flux subjected to the power normalization constraint. Verification of the LUNGA method demonstrated that the standard deviation in the number densities and infinite multiplication factor (kinf) predicted by this method agree well with the uncertainty obtained from the statistical analysis of 100 different simulations performed with coupled MC depletion calculations. Future research includes (a) expanding the LUNGA methodology to include more nuclides, (b) fully automating the methodology, and (c) investigating the use of an axial segmented fuel rod.
