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Nuclear Criticality Safety
NCSD provides communication among nuclear criticality safety professionals through the development of standards, the evolution of training methods and materials, the presentation of technical data and procedures, and the creation of specialty publications. In these ways, the division furthers the exchange of technical information on nuclear criticality safety with the ultimate goal of promoting the safe handling of fissionable materials outside reactors.
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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
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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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.
Chung-Yi Wang
Nuclear Technology | Volume 51 | Number 3 | December 1980 | Pages 332-348
Technical Paper | Mechanics Applications to Fast Breeder Reactor Safety / Reactor | doi.org/10.13182/NT80-A32571
Articles are hosted by Taylor and Francis Online.
An implicit finite difference method has been developed and incorporated into the ICECO code for analyzing hydrodynamics in the above-core region induced by the upper internal structure, and sodium spillage through penetrations and ruptured seals resulting from slug impact on the reactor cover. Eulerian description is employed so that flow through coolant passageways, large material distortions, two-dimensional sliding interfaces, flow around corners, and out-flow boundary conditions can be easily treated. In the analysis, the upper internals and the reactor cover are considered as perforated structures. A control-volume technique is utilized for deriving equations for the conservation of mass, momentum, and energy. The basic idea is to use actual fluid volume and actual flow areas in the mathematical formulation. Several modified Poisson equations are obtained, which govern the hydrodynamic pressures in the vicinity of the perforated structures. Sample problems are provided to illustrate the code capabilities in assessing the effect of the upper internal structure on the containment response and in estimating the amount of coolant ejected from the primary containment.