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Accelerator Applications
The division was organized to promote the advancement of knowledge of the use of particle accelerator technologies for nuclear and other applications. It focuses on production of neutrons and other particles, utilization of these particles for scientific or industrial purposes, such as the production or destruction of radionuclides significant to energy, medicine, defense or other endeavors, as well as imaging and diagnostics.
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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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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.
E. J. Allen, S. R. McNeany
Nuclear Technology | Volume 47 | Number 2 | February 1980 | Pages 363-377
Technical Paper | Analysis | doi.org/10.13182/NT80-A32439
Articles are hosted by Taylor and Francis Online.
Assay and analysis procedures were developed for nondestructive fissile isotopic measurement of mixed 233U-235U fuel samples. For 233U, the number of delayed neutrons released per fission is about half that for 235U, although the number of prompt neutrons is approximately the same. By separately counting prompt and delayed neutrons released by a sample exposed to neutron irradiation, the amounts of 233U and 235U present in the sample can be determined. Equations of delayed and prompt neutron counts versus 233U and 235U contents are solved simultaneously for the 233U and 235U contents of a sample. Eleven samples containing mixtures of 233U and 235U from no 233U to nearly 100% were prepared and assayed in prompt and delayed neutron assay devices. Constants for calibration equations were fitted to data from nine of the samples. The maximum differences between counts calculated by the calibration equations and measured counts were 2.3% for delayed neutrons and 1.2% for prompt neutrons, indicating a good selection of the form for the calibration equations. The two remaining samples were treated as unknown, and the uranium contents of these samples were determined by simultaneously solving the two calibration equations. The maximum difference between measured 233U or 235U content and actual content for either sample was 1.5%.
