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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
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.
Yasuo Koizumi, Hiroshige Kumamaru, Yuichi Mimura+, Yutaka Kukita, Kanji Tasaka†
Nuclear Technology | Volume 96 | Number 3 | December 1991 | Pages 290-301
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT91-A34590
Articles are hosted by Taylor and Francis Online.
Cold-leg small-break loss-of-coolant accident experiments were conducted for break areas ranging from 0.5 to 10% of the scaled cold-leg flow area using the Large-Scale Test Facility (LSTF). The LSTF is a volumetrically scaled simulator of a Westinghouse-type pressurized water reactor. For all the experiments, the core collapsed liquid level was temporarily depressed when liquid in the primary loop U-bend (crossover leg) was being cleared by steam. For scaled break areas <2.5%, the minimum core liquid level was equal to the lowest elevation of the crossover leg. For break areas >5%, the minimum core level was even lower because differential pressures created by the residual liquid holdup in the steam generator (SG) upflow side affected the core liquid level adversely. This influence of SG liquid holdup on the minimum core liquid level was larger for larger break sizes within the range of these experiments; thus, a more severe core level depression was seen for larger break sizes. Also, for the same break size, the core level depression was more severe when higher core power values were used for the simulation of the postscram core power decay. The RELAP5/MOD2 code reasonably well predicted the major phenomena observed in the experiments; however, several shortcomings were found in interfacial drag calculation for the SG U-tube inlet and the hot-leg outlet to the SG inlet plenum and core.