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Division Spotlight
Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
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.
Xue Zhou Jin
Fusion Science and Technology | Volume 77 | Number 5 | July 2021 | Pages 391-402
Technical Paper | doi.org/10.1080/15361055.2021.1904769
Articles are hosted by Taylor and Francis Online.
For the helium-cooled pebble bed breeding blanket concept improved in 2016 and the associated primary heat transfer system (PHTS) following EU DEMO Baseline 2015, an ex-vessel loss-of-coolant accident (LOCA) has been investigated with the assumption of a double-ended guillotine break of a main pipe in an outboard (OB) loop of the PHTS. The break leads to helium blowdown into the tokamak cooling room. A fast plasma shutdown followed by a plasma disruption is activated after the detection of the LOCA due to the design basis accident. Regarding three affected first-wall (FW) areas in one or two OB loops, three main cases are considered. If the FW temperature reaches the defined temperature limit of 1000°C, the FW is assumed to be failed such that an in-vessel LOCA results. In total five scenarios are simulated using MELCOR 1.8.6 for fusion with respect to the affected FW areas, mitigated or unmitigated plasma disruption conditions, the options of the dry or wet suppression tank, and the transport of source terms performed in the case of the beyond design basis accident without the plasma shutdown. The transient results are discussed for the time evolution of the accident sequences, pressurization in the systems, temperature behavior in volumes and structures, and tritium and dust transport behavior.