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Aerospace Nuclear Science & Technology
Organized to promote the advancement of knowledge in the use of nuclear science and technologies in the aerospace application. Specialized nuclear-based technologies and applications are needed to advance the state-of-the-art in aerospace design, engineering and operations to explore planetary bodies in our solar system and beyond, plus enhance the safety of air travel, especially high speed air travel. Areas of interest will include but are not limited to the creation of nuclear-based power and propulsion systems, multifunctional materials to protect humans and electronic components from atmospheric, space, and nuclear power system radiation, human factor strategies for the safety and reliable operation of nuclear power and propulsion plants by non-specialized personnel and more.
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.
Kun Mo, Hsiao-Ming Tung, Xiang Chen, Yang Zhao, Jon Hansen, James F. Stubbins
Nuclear Technology | Volume 183 | Number 3 | September 2013 | Pages 455-463
Technical Paper | Materials for Nuclear Systems | doi.org/10.13182/NT13-A19433
Articles are hosted by Taylor and Francis Online.
Both Alloy 617 and Alloy 230 have been considered the most promising structural materials for the Very High Temperature Reactor (VHTR). In this study, mechanical properties of both alloys were examined by performing tensile tests at three different strain rates and at temperatures up to 1000°C. This range covers time-dependent (plasticity) to time-independent (creep) deformations. Strain-rate sensitivity analysis for each alloy was conducted in order to approximate the long-term flow stresses. The strain-rate sensitivities for the 0.2% flow stress were found to be temperature independent (m [approximate] 0) at temperatures ranging from room temperature to 700°C due to dynamic strain aging. At elevated temperatures (800°C to 1000°C), the strain-rate sensitivity significantly increased (m > 0.1). Compared to Alloy 617, Alloy 230 displayed higher strain-rate sensitivities at high temperatures. This leads to lower estimated long-term flow stresses. Results of this analysis were used to evaluate the current American Society of Mechanical Engineers (ASME) allowable design limits for each alloy. The study showed that the allowable design stresses in the ASME Boiler and Pressure Vessel Code for either alloy do not provide adequate long-term degradation estimation. Nevertheless, rupture stresses for Alloy 617, developed in the ASME code case N-47-28, can generally satisfy the safety margins at 800°C and 1000°C estimated in the study following the strain-rate sensitivity analysis. Furthermore, additional material development studies might be required, since the design parameters for rupture stresses are constrained such that the current VHTR conceptual designs cannot satisfy the material limits.