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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.
P. E. Reagan, R. L. Beatty, E. L. Long, Jr.
Nuclear Science and Engineering | Volume 28 | Number 1 | April 1967 | Pages 34-41
Technical Paper | doi.org/10.13182/NSE67-A18664
Articles are hosted by Taylor and Francis Online.
Fuel particles coated with pyrolytic carbon are contemplated for use in several high-temperature gas-cooled reactors. This paper describes the performance of pyrolytic carbon-coated, high-density, uranium oxide particles irradiated at 1300 to 1600°C. The fission-gas release, burnups, and temperatures for five experiments are given. Coated particles with a builtin gap between the fuel and the inner laminar coating began to show evidence of failure by releasing bursts of fission gas after 27.9% uranium burnup, and postirradiation examination revealed delamination of the inner coating. Coated particles made with a porous carbon buffer layer between the fuel and an isotropic coating showed no evidence of failure by fission-gas release, and showed no damage due to irradiation when examined by metallography. Coated particles with neither gap nor buffer, but with a low-density inner coating applied directly to the fuel, retained fission gas successfully but showed enlargement of cracks that had formed at the fuel-coating interface during the coating process. The oxide particles did not flow at high burnup and expand into voids and cracks as the carbide particles did, and the oxide did not diffuse into the carbon coating at high temperatures.