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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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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.
HyeonTae Kim, Yonghee Kim
Nuclear Science and Engineering | Volume 195 | Number 5 | May 2021 | Pages 464-477
Technical Paper | doi.org/10.1080/00295639.2020.1839342
Articles are hosted by Taylor and Francis Online.
A thermomechanical fuel performance analysis module is implemented in the Korea Advanced Institute of Science and Technology Monte Carlo (MC) neutron transport code iMC. The module is designed particularly for advanced three-dimensional (3-D) fuel concepts, so an unstructured tetrahedral mesh grid is adopted for geometry flexibility. The cellwise detailed power density distribution is tallied from the MC transport and transferred to the thermomechanics module for the heat transfer, thermal expansion, and stress analysis. In this paper, a recently proposed 3-D fuel concept called the centrally shielded burnable absorber (CSBA) model was considered for numerical studies. Several fuel models were solved by the iMC code: a single CSBA pellet, a three-ball–loaded CSBA pellet, and a CSBA fuel-loaded 17 × 17 fuel assembly. From the analysis results, it was discovered that the uncertainty of the detailed power density distribution hardly affects the uncertainty of the thermomechanical analysis due to dissipation via conduction. Also, the importance of using detailed intrafuel power distribution data in such a thermal neutron spectrum has been demonstrated, showing about 30 K overestimation of peak temperature compared to the conventional uniform power assumption.