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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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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.
George R. Fegan, Daniel I. Herborn, Steven M. Lippincott
Nuclear Technology | Volume 37 | Number 1 | January 1978 | Pages 13-18
Technical Paper | Reactor | doi.org/10.13182/NT78-A32086
Articles are hosted by Taylor and Francis Online.
Received January 31, 1977 Accepted for Publication September 7, 1977 The net free volume of the containment is an essential parameter in the loss-of-coolant accident (LOCA) containment pressure analysis for pressurized water reactors. For an optimized emergency core cooling system performance due to the importance of backpressure during the reflood phase of a LOCA, it is necessary to have the predicted pressure quite close to the design pressure. Using a geometric analysis, an estimate of 56 241.99 m3 (1 986 167 ft3) for the net free volume has been made for the containment of the Trojan nuclear plant. Two sets of data were produced from the normally scheduled structural integrity and integrated leak-rate tests on the Trojan containment. These data sets were used to arrive at two new estimates of the net free volume. A deterministic equation giving volume as a function of the slope of a linear relationship between depressurization and time was developed. After an analysis of the reliability of the data, estimates of this linear slope were made from the two data sets. These two slopes gave net free volume estimates of 58 000 m3 (2.05 × 106 ft3) and 57 650 m3 (2.036 × 106 ft3) when used in the deterministic equation. The maximum deviation from the geometric-based estimate was <4%.
