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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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Nuclear Science and Engineering
June 2025
Nuclear Technology
Fusion Science and Technology
May 2025
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.
X. Gaus-Liu, A. Miassoedov, J. Foit, T. Cron, F. Kretzschmar, Alexander Palagin, T. Wenz, S. Schmidt-Stiefel
Nuclear Technology | Volume 181 | Number 1 | January 2013 | Pages 216-226
Technical Paper | Special Issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) / Fission Reactors; Reactor Safety | doi.org/10.13182/NT13-A15769
Articles are hosted by Taylor and Francis Online.
The LIVE-L4 and LIVE-L5L experiments investigated the thermal-hydraulic behavior of the corium pool in the reactor pressure vessel lower head with the three-dimensional test vessel LIVE. The simulant material is a noneutectic binary mixture of 20% NaNO3-80% KNO3. Transient and steady-state parameters such as melt temperature and heat flux distribution through the vessel wall as well as crust formation characteristics were obtained. The two tests demonstrated that transient events like repeated melt relocation and change of decay power density facilitate crust deformation and change of crust thickness. Massive crust formation in a noneutectic melt pool leads to a change of melt pool composition and a decrease of melt-crust interface temperature. The melt temperature and heat flux at the same pool height and same power density can be roughly compared independent of heating history and initial melt pouring pattern. The dimensionless melt temperature as well as the dimensionless heat flux through the wall during the steady state are independent of power density if the pools have the same height. But, they are dependent on the pool height. For a low pool, the gradients with height of both melt temperature and heat flux through the vessel are larger than those for a high pool.
