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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
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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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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.
J. H. Shaffer, W. R. Grimes, G. M. Watson, D. R. Cuneo, J. E. Strain, M. J. Kelly
Nuclear Science and Engineering | Volume 18 | Number 2 | February 1964 | Pages 177-181
Technical Paper | doi.org/10.13182/NSE64-A18316
Articles are hosted by Taylor and Francis Online.
In the conceptual two-region molten-salt breeder reactor, fissionable U233 will be recovered from the blanket as the decay product of Pa233. Since equilibrium concentrations of Pa233 would result in appreciable parasitic neutron absorptions, the advantages of thermal breeding could be realized to a greater extent by removing both Pa233 and U233 from the blanket mixture. Methods for recovering these materials from molten-fluoride mixtures by precipitation as oxides are presented. Small-scale experiments clearly indicated that it is possible to remove protactinium from molten-fluoride solutions by a process that appears to be surface precipitation of protactinium on beryllium oxide or thorium oxide particles. Protactinium was removed from molten mixtures of LiF-BeF2-ThF4 (67-18-15 mole %) by the addition of 1 to 2% by weight of solid beryllium oxide or thorium oxide. The removal efficiency was high when the initial concentration of protactinium was either in the range 1 to 2 ppb or 50 to 75 ppm. Uranium was successfully removed from solution in molten fluorides by use of a similar procedure. Approximately 2000 ppm uranium was precipitated from molten LiF-BeF2-ThF4 (67-18-15 mole %) by the addition of 3% by weight of beryllium oxide. Comparable results were also obtained using thorium oxide as the precipitant.
