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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.
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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
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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.
Kyra Lawson, Nicholas R. Brown
Nuclear Technology | Volume 210 | Number 11 | November 2024 | Pages 2133-2150
Research Article | doi.org/10.1080/00295450.2024.2310911
Articles are hosted by Taylor and Francis Online.
This work presents neutronics models of a small and large fast-spectrum molten chloride-salt reactor. The models are similar to designs being pursued by industry, and they may serve as generic preconceptual and simplified neutronics models that provide information for decision making in licensing-related areas. The two models were created using Serpent, a Monte Carlo neutron transport code, and Moltres, a neutron diffusion core simulator tool. Specifically, this study focused on exploring the applicability of diffusion theory to fast molten salt reactor (MSR) models, the capabilities of an open-source, MSR-oriented simulation tool (Moltres), and optimal energy-group structures.
The proposed two-step method involves group-constant generation with Serpent and a multigroup diffusion solution by Moltres. Three energy-group structures were applied. The accuracy of the solutions was determined through comparisons between the two-step and Monte Carlo flux and multiplication factor solutions.
The findings indicated diffusion theory captures neutronics with minimal error for the large MSR and yielded best results with the 27-group structure. The 27-group structure yielded an average group flux error below 2% and keff agreement between diffusion and transport solutions within 30 pcm. The accuracy of the two-step method decreased for the very small (high-leakage) fast chloride MSR, but the neutronics were captured acceptably well with the 33-group structure.
In addition to exploring the capabilities of Moltres, this work contributes to the sparse literature involving open-source models of fast-spectrum MSRs. Future work is noted as expanding the capabilities of the neutronics models to incorporate thermal hydraulics.
