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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.
F. Payot, C. Journeau, C. Suteau, F. Serre (CEA), M. Gradeck, N. Rimbert, A. Lecoanet (Univ of Lorraine), A. Miassoedov (KIT)
Proceedings | 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) | Charlotte, NC, April 8-11, 2018 | Pages 657-666
The strategy used in ASTRID Sodium Fast Reactor Demonstrator to mitigate the consequences of a postulated severe accident relies on the presence of dedicated corium discharge tubes between the active core region and the lower plenum and an in-vessel core catcher in the lower plenum to collect and cool the molten fuel. In this situation, the corium progression from the core towards the core catcher could lead to jet impingement on the core catcher surface and a subsequent degradation of the core-catcher material(s). After a review of the existing experimental database, it appeared necessary to define a dedicated experimental R&D program related to the long-lasting jet impingement on thick material plates. A particular behaviour will be studied when a molten pool is created (named “pool effect”) at the impingement point that reduces the heat transfer at the jet-material interface. In this scope, experimental tests with simulants and prototypic materials will be carried out by using the most possible representative conditions (e.g. Reynolds and Prandtl numbers) for ESFR (European Sodium Fast Reactor) severe accident conditions. Three R&D program have been proposed:
(i) Ice-water jet impingement, the few results in the literature show a good agreement between ice-water jet system and metal wall/metal jet system which justifies the use of such simulant materials for investigating the impingement behaviour. The JOLO facility will be designed at LEMTA Nancy University.
(ii) Molten steel jet impingement on the thick wall. The MOCKA facility in the Karlsruhe Institute of Technology will be used.
(iii) Dedicated test section PLINIUS-2 IMPACT will be designed in future PLINIUS 2 large-mass prototypic-corium experimental platform at CEA Cadarache.
