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Division Spotlight
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.
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.
Motoo Aoyama, Sadao Uchikawa, Kazuyoshi Miki, Kazuo Hiramoto, Renzo Takeda
Nuclear Technology | Volume 64 | Number 1 | January 1984 | Pages 19-25
Technical Paper | Nuclear Fuel | doi.org/10.13182/NT84-A33323
Articles are hosted by Taylor and Francis Online.
A new design concept of a boiling water reactor (BWR) fuel bundle for extended burnup was proposed to improve the capacity factor without increasing the fuel cycle cost. Some effects, which are raised from higher burnup, such as strong pellet-cladding interaction due to enhanced fuel swelling and changes in neutronic characteristics due to increased fuel enrichment, are minimized by a reduction in the maximum fuel temperature to below 1200°C and an increase in the moderator-to-fuel ratio. To realize these concepts, a 9 × 9 lattice design with a reduced fuel rod diameter and annular pellets was proposed. The proposed fuel bundle design offers advantages in fuel cycle improvements through extension of achievable burnup and reduction of fuel inventory. The core, loaded with the proposed fuel bundles which achieve 30% higher burnup by the full power month, has a potential for natural uranium savings of ∼20% per unit power and a reduction in the amount of reprocessing of ∼40% per unit power, compared with the current BWR design when coupled with other improvements such as refueling pattern optimization, natural uranium axial blankets, and spectral shift with flow control.