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Division Spotlight
Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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.
Canhui Sun, Zhaocan Meng, Yaodong Chen (State Power Investment Corp. Research Inst)
Proceedings | 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) | Charlotte, NC, April 8-11, 2018 | Pages 1153-1160
Rapid industrialization and urbanization in the last 30 years in China have caused severe air pollution. In order to meet the new challenges on the energy and environment, a new conceptual design of nuclear heating reactor (HAPPY200) is developed. HAPPY200 is a two-loop low pressure water reactor with low thermal parameters and passive safety systems. It can be used for heating, cooling, desalination and other process heat applications. Based on large volume pool completely passive technologies, HAPPY200 safety system can provide an inherent technical guarantee for the safety of the reactor.
The thermal-hydraulic behavior analysis is necessary for the safety and economy of the design. Subchannel analysis is the basic thermal-hydraulic analysis method used to predict the coolant enthalpy, quality, density, mass velocity rate, liquid temperature, vapor void fraction, pressure distribution, and the resulting DNBR distributions.
In this study, the analysis of thermal-hydraulic behavior for low pressure reactor is worked. Some problems are discussed for low pressure reactor, and the subchannel object model is established by subchannel code. Some low pressure CHF correlations developed for other fuel assemblies are analyzed. The thermal diffusion coefficient for reduced-height rod bundle is obtained compared with full-height rod bundle using CFD code. The value of reduced-height rod bundle can be used in the preliminary analysis of HAPPY200. According to the subchannel results, all the values satisfy conceptual design criteria.
