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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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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.
Kazuyuki Takase, Yasuo Ose, Hajime Akimoto
Fusion Science and Technology | Volume 39 | Number 2 | March 2001 | Pages 1050-1055
Safety and Environment | doi.org/10.13182/FST01-A11963382
Articles are hosted by Taylor and Francis Online.
Damage of cooling tubes of plasma facing components (PFCs) results in water discharge into a vacuum vessel (W) of a fusion reactor. Flashing in vacuum, water pool boiling and impingement-jet on a surface of the PFC are the main heat transfer phenomena responsible for steam production that causes a rapid pressurization of the W. This is called an in-vessel loss-of-coolant accident (LOCA) event or ingress-of-coolant event (ICE). The ICE event is one of the most severe accidents in the fusion reactors.
The integrated ICE test facility was constructed to demonstrate the safety design approach of International Thermonuclear Experimental Reactor (ITER) and obtain validation data for the ITER safety analysis codes. Then, an experimental study was performed using the integrated ICE test facility and at the same time the code validation study with the TRAC code was carried out. The pressure rise characteristics in the current ITER machine during the ICE event were analyzed numerically using the verified TRAC-PF1 code and the effects of the relief pipe diameter and suppression tank volume regarding to the pressure rise due to the ICE events were clarified quantitatively from the present analytical results.