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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
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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Nuclear Science and Engineering
June 2025
Nuclear Technology
Fusion Science and Technology
May 2025
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.
Muhammad Ishaq, Muhammad Zaman, Muhammad Ilyas, Alam Nawaz Khan Wardag, Mansoor H. Inayat
Nuclear Science and Engineering | Volume 198 | Number 12 | December 2024 | Pages 2382-2402
Research Article | doi.org/10.1080/00295639.2024.2328967
Articles are hosted by Taylor and Francis Online.
Innovative reactor designs like small modular reactors (SMRs) have the potential to operate in a natural circulation (NC) boiling mode, but this mode introduces flow oscillations that pose a risk to nuclear safety. Therefore, it is essential to investigate the effects of various parameters on these oscillations. This study focuses on predicting the operational behavior of the Integral PWR-type SMR Test Rig (iPSTR) when operating in NC and subcooled boiling conditions. The iPSTR replicates an NC boiling loop with a vertical heater, vertical cooler configuration, high-temperature and high-pressure conditions, and nonuniform diameter structure. Using the RELAP5 model, thermal-hydraulic simulations were performed to anticipate how varying degrees of inlet subcooling affects parameters such as mass flow rate and void fraction, with experimental data used to validate the model’s accuracy. This investigation covers a range of process conditions, including system pressures from 5 to 20 bars, core input power varying from 8.5 to 14.5 kW, and degrees of inlet subcooling from 1 to 49 K. The results reveal that increasing input power leads to higher average mass flow rates, while at a constant system pressure, higher input power stabilizes flow rates at higher degrees of inlet subcooling. Moreover, reduced and more consistent oscillation amplitudes and frequencies at higher core power result at more elevated system pressure, enhancing the safety of the iPSTR facility.