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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.
Rei Kimura, Kazuhito Asano
Nuclear Technology | Volume 210 | Number 8 | August 2024 | Pages 1496-1502
Note | doi.org/10.1080/00295450.2023.2299899
Articles are hosted by Taylor and Francis Online.
A novel microreactor, called MoveluXTM, was previously proposed that utilizes heat pipes as the primary heat transfer device and calcium hydride as the moderator. In this core design, the moderator temperature is the critical core operation limit because at high temperatures above 800°C, the hydrogen dissociates from the calcium hydride. The core temperature distribution, therefore, was previously evaluated. However, this evaluation did not consider gamma heating in the core and assumed that power was produced only in the fuel region. By contrast, the moderator region has a power density under realistic conditions due to gamma heating. Thus, the present work considers gamma heating in the core power distribution calculation and evaluates the impact on the moderator temperature. The power density of gamma heating was 1/10th that of the fuel region and around 1/100th that of the core thermal power. This increased the temperature of the moderator by 10 K from the case without considering gamma heating. In addition, this temperature distribution difference did not have an impact on the core criticality. In conclusion, considering the gamma heating, concerns regarding the core design are not suggested.