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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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.
Werner Schenk, Heinz Nabielek
Nuclear Technology | Volume 96 | Number 3 | December 1991 | Pages 323-336
Technical Paper | Nuclear Fuel Cycle | doi.org/10.13182/NT96-3-323
Articles are hosted by Taylor and Francis Online.
The essential feature of small, modular high-temperature reactors (HTRs) is the inherent limitation in maximum accident temperature to below 1600°C combined with the ability of coated particle fuel to retain all safety-relevant fission products under these conditions. To demonstrate this ability, spherical fuel elements with modern TRISO particles are irradiated and subjected to heating tests. Even after extended heating times at 1600°C, fission product release does not exceed the already low values projected for normal operating conditions. Details of fission product distribution within spherical fuel elements heated at constant temperatures of 1600, 1700, and 1800°C are presented. The measurements confirm the silicon carbide (SiC) coating layer as the most important fission product barrier up to 1800° C. If the SiC fails (or is defective), the following transport properties at 1600 to 1800°C can be observed: cesium shows the fastest release from the UO2 kernel but is highly sorbed in the buffer layer of the particle and in the matrix graphite of the sphere; strontium is retained strongly both in UO2 kernels and in matrix graphite, but can penetrate SiC in some cases where cesium is still completely retained; only if all coating layers are breached can iodine and noble gases be released. For the first 100 h at 1600°C (enveloping all possible accident scenarios of small HTRs), these fission products are almost completely retained in the coated particles.