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Division Spotlight
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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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.
Peter Yarsky
Nuclear Technology | Volume 207 | Number 5 | May 2021 | Pages 665-679
Technical Paper | doi.org/10.1080/00295450.2020.1810465
Articles are hosted by Taylor and Francis Online.
In a companion paper, the U.S. Nuclear Regulatory Commission (NRC) staff has described analyses performed using the TRAC/RELAP Advanced Computational Engine (TRACE) code to study the transient system response of the NuScale power module to a postulated beyond-design-basis loss of alternating-current (LOAC) power transient where the module protection system completely fails to insert the control rods. The subject paper studies the sensitivity of the event progression and consequences to variation in the initial reactor coolant system (RCS) temperature. These studies were performed by varying the effective steam generator heat transfer surface area between 100% and 50% of the nominal area. The results of the NRC staff analyses show that at increased initial temperatures, it is possible for the NuScale primary side to remain critical for an extended period of time, leading to a sustained loss of primary-side inventory through pressure relief until the natural circulation flow pattern in the RCS becomes broken. After the flow loop is broken, reactor power decreases significantly, and the primary figures of merit important to safety are met with substantial margin.