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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.
Hyo-Nam Kim, Ihn Namgung
Nuclear Technology | Volume 195 | Number 1 | July 2016 | Pages 15-28
Technical Paper | doi.org/10.13182/NT15-17
Articles are hosted by Taylor and Francis Online.
In severe accident conditions, the molten core material forms an internally heated debris bed and eventually becomes a molten pool of corium, which will cause or induce thermal and mechanical loads to the reactor vessel lower head (RVLH) resulting in penetrations leading to failure. A good understanding of the mechanical behavior of the RVLH is essential for estimating structural integrity and improving accident mitigation strategies.
Coupled thermomechanical analysis using ANSYS, a general-purpose finite element method analysis code, was used to evaluate the possibility and timescale of failure. A two-dimensional axisymmetric finite element model was adopted based on APR1400 design data with relevant material properties including creep data.
From the study, it was found that the possibility of plastic and creep failure of the RVLH for the APR1400 was considerably low for a full-core meltdown of the reactor core under ex-vessel cooling conditions with an ambient temperature of 130°C and constant decay heat from the corium, but the lower head may fail unless the increased internal pressure can be reduced on time. Plastic failure can be a major cause of lower head failure of a reactor vessel in high internal pressure conditions and creep failure is not negligible, since failure mechanisms under long-lasting periods are considered. This study found that the APR1400 RVLH failure time is around 220 h using 15% creep strain failure criteria from the postulated accident condition.