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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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.
Seok-Hee Ryu, Kil-Sup Um, Jae-Il Lee
Nuclear Technology | Volume 189 | Number 2 | February 2015 | Pages 163-172
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT14-28
Articles are hosted by Taylor and Francis Online.
To evaluate the effect of thermal conductivity degradation for high-burnup fuel, a postulated control element assembly (CEA) ejection accident is assessed with the SPACE (Safety and Performance Analysis CodE) code. The SPACE code, which is currently under development as a safety analysis code for nuclear power plants, can predict thermal-hydraulic responses of the nuclear fuel and nuclear steam supply system during design basis accidents with two-fluid, three-field governing equations. Fuel performance behaviors during the CEA ejection accident using six fuel conductivity models including the burnup-independent reference conductivity model, the Lyons model, are investigated and compared with results of the reference model within the range from 0 to 30 GWd/tonne U. The Oak Ridge National Laboratory model predicts the highest peak fuel centerline temperature of 4531°F at 0 GWd/tonne U, and the modified Nuclear Fuels Institute model shows the uppermost value of 4796°F, which is 364°F higher than the reference model at 30 GWd/tonne U. It is also observed that the peak fuel centerline temperature increases linearly with fuel burnup and that the maximum increase rate of the peak centerline temperature per fuel burnup is ∼11.6°F per GWd/tonne U. For all thermal conductivity models, the maximum radial average fuel enthalpies are <230 cal/g, and the rise in radial average fuel enthalpy during the CEA ejection accident still remains within the pellet-cladding-mechanical-interaction failure criterion.