ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
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Division Spotlight
Accelerator Applications
The division was organized to promote the advancement of knowledge of the use of particle accelerator technologies for nuclear and other applications. It focuses on production of neutrons and other particles, utilization of these particles for scientific or industrial purposes, such as the production or destruction of radionuclides significant to energy, medicine, defense or other endeavors, as well as imaging and diagnostics.
Meeting Spotlight
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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Reviewers needed for NRC research proposals
The deadline is fast approaching for submitting an application to become a technical reviewer for the Nuclear Regulatory Commission’s fiscal year 2025 research grant proposals.
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.
