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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.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott 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
Smarter waste strategies: Helping deliver on the promise of advanced nuclear
At COP28, held in Dubai in 2023, a clear consensus emerged: Nuclear energy must be a cornerstone of the global clean energy transition. With electricity demand projected to soar as we decarbonize not just power but also industry, transport, and heat, the case for new nuclear is compelling. More than 20 countries committed to tripling global nuclear capacity by 2050. In the United States alone, the Department of Energy forecasts that the country’s current nuclear capacity could more than triple, adding 200 GW of new nuclear to the existing 95 GW by mid-century.
Jorge Solís, Maria N. Avramova, Kostadin N. Ivanov
Nuclear Technology | Volume 146 | Number 3 | June 2004 | Pages 267-278
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT04-A3505
Articles are hosted by Taylor and Francis Online.
A multilevel methodology has been developed to extend the TRAC-BF1/NEM coupled code capability to obtain the transient fuel rod response. The COBRA-TF thermal-hydraulics subchannel analysis code is coupled to TRAC-BF1/NEM in the parallel virtual machine environment. The power information obtained from the nodal expansion method three-dimensional neutronic calculation is used by the hot subchannel analysis module. The TRAC-BF1 thermal-hydraulic system analysis code provides the COBRA-TF thermal-hydraulic boundary conditions. The subchannel analysis module uses this information to recalculate the fluid, thermal, and hydraulics conditions in the most limiting node (axial region of assembly/channel) within the core at each time step. A dynamic algorithm has been developed to identify the most limiting channel and fuel assembly (radially) and axial region (node) based on the current state of the core. Results, obtained with the new parallel multilevel coupled methodology, are presented and discussed for the Mexican Laguna Verde 1 nuclear power plant control rod drop accident.
