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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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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.
H.Yoshida, M.Taniguchi, K.Yokoyama, Y.Hirohata, M.Akiba, T.Hino
Fusion Science and Technology | Volume 41 | Number 3 | May 2002 | Pages 943-947
Material Interaction and Permeation | Proceedings of the Sixth International Conference on Tritium Science and Technology Tsukuba, Japan November 12-16, 2001 | doi.org/10.13182/FST02-A22724
Articles are hosted by Taylor and Francis Online.
Tritium retention of carbon dust co-deposited with fuel hydrogen is large, and then it is required to evaluate the tritium inventory as a safety issue of ITER. Several species of co-deposited carbon dust were prepared by D2 arc discharge with carbon electrodes. The dependence of D2 gas pressure on the retained deuterium amount of the co-deposited dust was investigated. The structure and the surface morphology were also examined. The retained deuterium amount increased with the discharge gas pressure. The deuterium concentrations of the co-deposited carbon dust samples prepared at 1.3 Pa and 6.8 Pa were 0.12 and 0.3 in the atomic ratio, D/C, respectively. No clear dependence of the substrate temperature on retained deuterium amount was observed, perhaps due to the coarse temperature control. In the environment of gas pressure in ITER, approximately 1 Pa, the tritium concentration is estimated approximately T/C ≈ 0.06, which is several times smaller than the value estimated so far, T/C ≈ 0.2.