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Division Spotlight
Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
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.
Hyungrae Kim, Yoon Yeong Bae, Hwan Yeol Kim, Jin Ho Song, Bong Hyun Cho
Nuclear Technology | Volume 164 | Number 1 | October 2008 | Pages 119-129
Technical Paper | Icapp '06 | doi.org/10.13182/NT08-A4013
Articles are hosted by Taylor and Francis Online.
The SuperCritical Water-cooled Reactor (SCWR) is one of the candidates for the fourth-generation nuclear power plant, and it uses light water as a coolant. Heat transfer between a fuel assembly and water may degrade at certain conditions of supercritical pressure flows. Therefore, accurate and reliable estimation of heat transfer coefficients is necessary for the design of the fuel assembly and the reactor core. A series of heat transfer tests has been carried out at a test facility named SPHINX by using carbon dioxide as a stimulant of water. The tests produced heat transfer data of the supercritical pressure flows inside a circular tube of 4.4-mm inside diameter at varying operating pressures, mass fluxes, and wall heat fluxes. The test range of the mass flux was 400 to 1200 kg/m2 s, and the maximum heat flux was 150 kW/m2. The operating pressures were 7.75, 8.12, and 8.85 MPa in the tests. The test results were compared with estimations of the existing correlations for supercritical pressure flows. The comparison showed reasonable agreement between our data and the correlations. However, a rather large departure from the normal heat transfer correlations was observed near pseudocritical temperatures. Besides the comparison of the normal heat transfer coefficients, the onset of heat transfer deterioration was compared between the test cases and two existing criteria. One of the criteria was derived from experiments by using Freon but with a test section of identical geometry while the other criterion was for a flow of carbon dioxide in a larger bore channel than ours. Both criteria showed fair agreement with our experiment. Most test cases with noticeable heat transfer degradation were located in the region of deterioration predicted by the criteria.