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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.
Hsiang-Shou Cheng, David J. Diamond
Nuclear Technology | Volume 56 | Number 1 | January 1982 | Pages 40-54
Technical Paper | Fission Reactor | doi.org/10.13182/NT82-A32879
Articles are hosted by Taylor and Francis Online.
The center rod drop accident was calculated for a boiling water reactor using the two-dimensional (R,Z) core dynamics code BNL-TWIGL. Analysts frequently neglect moderator feedback under the assumption that it leads to conservative results. The present study shows that the peak of the power burst and peak fuel enthalpy can indeed be reduced by a factor of 2 or more by including this effect. The magnitude of the effect depends on reactor conditions. Moderator feedback is particularly important when there are voids in the core initially (i.e., at power conditions) or when the core is near saturation condition. When the reactor is initially at zero power and considerably subcooled, moderator feedback will influence the power peak by <10% but will have a much larger effect on the peak fuel enthalpy, which occurs later in time. The moderator feedback is the result of heat conducted from the fuel rod and direct energy deposition. At power conditions, the time constant for heat conduction is small and this is the primary mechanism for changing the steam void content during the accident. At zero power, the initial thermal constant is very large and, hence, any generation of voids at short times is due to direct energy deposition in the moderator. The effect of a different initial power level, flow rate, and inlet sub cooling, as well as the effect of delayed neutron fraction, rod drop speed, and accident rod worth, was calculated. In all cases, with moderator feedback accounted for, the maximum fuel enthalpy during the accident is well below presently established limits. Accident consequences are insensitive to the delayed neutron fraction and rod drop velocity. The parameters of most significance are inlet subcooling and accident rod worth. Most of the analysis used a fixed inlet flow and core pressure. A plant transient calculation was run to see how these parameters varied. The result was fed back into a bounding core calculation, which then showed that the change in pressure and flow increases the peak fuel enthalpy but not to an appreciable extent.