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Nuclear Installations Safety
Devoted specifically to the safety of nuclear installations and the health and safety of the public, this division seeks a better understanding of the role of safety in the design, construction and operation of nuclear installation facilities. The division also promotes engineering and scientific technology advancement associated with the safety of such facilities.
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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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.
Russell D. Mosteller, Peter J. Jensen, Michael J. Anderson,+ Laurance D. Eisenhart, Rana Abdollahian, Jason Chao, Walter J. Eich
Nuclear Technology | Volume 86 | Number 1 | July 1989 | Pages 40-48
Technical Paper | Nuclear Safety | doi.org/10.13182/NT89-A34280
Articles are hosted by Taylor and Francis Online.
A pressurized water reactor (PWR) with a positive moderator temperature coefficient of reactivity is potentially susceptible to a severe overheating transient. This study identifies a scenario in which such a transient could occur and is similar in some respects to the accident at Chernobyl Unit 4. The scenario so identified is a natural circulation test at beginning of life under the assumption that all scrams are disabled. The results obtained demonstrate that a runaway power excursion does not occur and that the domestically designed PWR that was analyzed displays inherently safe behavior for the chosen scenario. The analysis is performed using two codes in tandem over three sequential stages of the analysis. Since the transient is assumed to be quasi-steady-state, steadystate calculations first are performed with the advanced three-dimensional core simulation code ARROTTA to generate a moderator reactivity table for input to the point kinetics model in the RETRAN-02 systems transient code. The entries in this table are taken directly from ARROTTA results, so they implicitly contain three-dimensional reactivity effects. Next, a RETRAN-02 transient analysis is performed for the system as a whole. This analysis serves two purposes: It predicts the overall behavior of the plant during the transient, and it also produces time-dependent forcing functions for ARROTTA. Finally, ARROTTA is run in a transient mode, providing a detailed description of the behavior of the core during the transient. The high level of consistency produced by the two transient calculations validates the initial assumption of quasisteady-state behavior. A methodology that may be applied to the analysis of this and similar transients is developed and tested as part of this study.