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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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ANS designates Armour Research Foundation Reactor as Nuclear Historic Landmark
The American Nuclear Society presented the Illinois Institute of Technology with a plaque last week to officially designate the Armour Research Foundation Reactor a Nuclear Historic Landmark, following the Society’s decision to confer the status onto the reactor in September 2024.
Ian J. Hastings, Elio Mizzan, Alan M. Ross, John R. Kelm, Real J. Chenier, D. H. Rose, J. Novak
Nuclear Technology | Volume 68 | Number 1 | January 1985 | Pages 40-47
Technical Paper | Nuclear Fuel | doi.org/10.13182/NT85-A33565
Articles are hosted by Taylor and Francis Online.
Fragments of UO2 fuel pellets extracted from irradiated elements were heated in air at 175 to 275 °C for times up to 800 h. Unirradiated pellets and fragments were studied for comparison. Pretest burnup of the irradiated fuel was typically 190 MW-h/kgU (7900 MWd per tonne of uranium) at a maximum linear power of 45 kW/m. The fuel had been discharged for 1 to 3 yr. The maximum weight gain was at 275 °C, ∼4% in 70 h, indicating 100% conversion to U3O8. The activation energy for the oxidation process at 175 to 275 °C was 130 ± 10 kJ/mol. There was a strong effect of prior irradiation on oxidation rate; the weight gain at 250 °C was about a factor of 6 greater in irradiated compared with unirradiated fuel. There was also an effect of fragment size on oxidation rate. Also, weight gains of fragments from a naturally defected element were less than those for fragments from intact fuel, consistent with prior oxidation in the defected state.