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Division Spotlight
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.
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.
Young Min Kim, Moon Sung Cho
Nuclear Technology | Volume 170 | Number 1 | April 2010 | Pages 231-243
Technical Paper | Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fuel Cycle and Management | doi.org/10.13182/NT10-A9461
Articles are hosted by Taylor and Francis Online.
The COPA-FPREL computer code has been developed to estimate the releases of gaseous and metallic fission products (FPs) from high-temperature gas-cooled reactor (HTGR) fuel into coolant. The COPA-FPREL code treats FP release from a coated fuel particle (CFP), diffusion in a fuel element, and leakage into the coolant considering the temperature distribution within a CFP and a fuel element. The code uses a finite difference method to calculate FP migration and heat transfer. In the finite difference method, the kernel, buffer, and coating layers of a CFP and the fuel element are divided into small finite difference intervals. A steady-state heat transfer equation and the Fickian diffusion equation are applied to these intervals. A relatively high diffusion coefficient is assigned to the buffer and the broken coating layers to describe fast diffusion in those regions. Sorption equilibrium is set up between the concentration at the fuel element surface facing the coolant and the vapor pressure at the graphite side of the boundary layer that forms on the fuel element surface. Mass transfer occurs through the boundary layer into the bulk coolant. In a prismatic HTGR, sorption equilibrium is assumed to form between the concentrations at the compact and structural graphite surfaces and the vapor pressure in the gap between the compact and the structural graphite. For 137Cs, 90Sr, 110mAg, and 85Kr isotopes, the fractional releases from a CFP, a pebble, and a fuel block during simulated heating processes and reactor operations were calculated using COPA-FPREL.