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Fuel Cycle & Waste Management
Devoted to all aspects of the nuclear fuel cycle including waste management, worldwide. Division specific areas of interest and involvement include uranium conversion and enrichment; fuel fabrication, management (in-core and ex-core) and recycle; transportation; safeguards; high-level, low-level and mixed waste management and disposal; public policy and program management; decontamination and decommissioning environmental restoration; and excess weapons materials disposition.
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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
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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.
M. Zabiégo, F. Fichot, P. Rubiolo
Nuclear Technology | Volume 154 | Number 2 | May 2006 | Pages 194-214
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT06-A3728
Articles are hosted by Taylor and Francis Online.
In the frame of Institut de Radioprotection et de Sûreté Nucléaire research programs on severe accidents in pressurized water reactors (PWRs), a new radiative heat transfer model to be used in the ICARE/CATHARE software is presented. The reactor core is considered an optically thick porous medium, and the diffusion approximation is adopted. The equivalent conductivity of the medium is determined. Its expression is carefully established to take into account the strong geometrical variations occurring in a reactor core undergoing a severe accident sequence (as observed in Three Mile Island Unit 2). After describing the theoretical basis of our approach, it is shown that the continuity of the equivalent conductivity is ensured when the geometry evolves from an array of intact cylinders to a particle bed.When compared to the more classical radiation method used in most severe accident codes, this approach better predicts the radial temperature gradient obtained by Cox in his experiment in bundle geometry. The same comparison on a PWR vessel undergoing an accidental sequence brings to the fore the impact of the radiation modeling on the degradation process: The sideward heat losses predicted by the method selected in this work are more limited, which slows the radial progression of the degradation.