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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.
Rafael Macian, Paul Coddington
Nuclear Technology | Volume 139 | Number 3 | September 2002 | Pages 185-204
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT02-A3313
Articles are hosted by Taylor and Francis Online.
RETRAN-3D, a system analysis code currently employed by the nuclear industry in studies covering a wide variety of operational and accident scenarios, has not been extensively validated for application to loss-of-coolant accident (LOCA) scenarios.The results of the in-depth analysis of two experimental loss-of-coolant transients, namely, Test No. 9 in the French OMEGA facility, and the International Standard Problem 26 (ISP-26) in the Japanese ROSA-IV Facility are discussed. The OMEGA test simulated the blowdown phase of a double-ended cold-leg break, whereas the ISP-26 test simulated a small break (5%) in a full height, volume (1/48), and power (~1/342) scaled facility representing a typical two (or four)-loop pressurized water reactor (PWR) system.The RETRAN-3D results for the OMEGA test show good estimates of the important system parameters, with the best agreement corresponding to the use of the dynamic-slip flow model. A sensitivity analysis on the break flow showed that the Henry/Fauske-Isoenthalpic Expansion critical flow model yields the best results, which are significantly improved with a refined nodalization upstream of the break.The ISP-26 was also simulated using the dynamic-slip flow model. The results indicate that the code is able to calculate a small-break LOCA with a model including the main PWR system components and to reproduce the principal physical processes in a reasonable manner.In summary, this assessment shows the ability of RETRAN-3D to model LOCA scenarios in a reasonable way and also points to areas where further model improvement could result in more accurate simulations.
