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Division Spotlight
Thermal Hydraulics
The division provides a forum for focused technical dialogue on thermal hydraulic technology in the nuclear industry. Specifically, this will include heat transfer and fluid mechanics involved in the utilization of nuclear energy. It is intended to attract the highest quality of theoretical and experimental work to ANS, including research on basic phenomena and application to nuclear system design.
Meeting Spotlight
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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High-temperature plumbing and advanced reactors
The use of nuclear fission power and its role in impacting climate change is hotly debated. Fission advocates argue that short-term solutions would involve the rapid deployment of Gen III+ nuclear reactors, like Vogtle-3 and -4, while long-term climate change impact would rely on the creation and implementation of Gen IV reactors, “inherently safe” reactors that use passive laws of physics and chemistry rather than active controls such as valves and pumps to operate safely. While Gen IV reactors vary in many ways, one thing unites nearly all of them: the use of exotic, high-temperature coolants. These fluids, like molten salts and liquid metals, can enable reactor engineers to design much safer nuclear reactors—ultimately because the boiling point of each fluid is extremely high. Fluids that remain liquid over large temperature ranges can provide good heat transfer through many demanding conditions, all with minimal pressurization. Although the most apparent use for these fluids is advanced fission power, they have the potential to be applied to other power generation sources such as fusion, thermal storage, solar, or high-temperature process heat.1–3
Taehwan Ahn, Julio Diaz, Robert Adams, Victor Petrov, Annalisa Manera
Nuclear Technology | Volume 209 | Number 12 | December 2023 | Pages 1898-1913
Research Article | doi.org/10.1080/00295450.2023.2197680
Articles are hosted by Taylor and Francis Online.
High-resolution two-phase flow data in the rod bundle are important in the development and validation of high-fidelity models for computational fluid dynamics and subchannel codes, in particular, those pertaining to light water reactor cooling systems. The Michigan Adiabatic Rod Bundle Flow Experiment (MARBLE) has been constructed as a modular assembly of an 8 × 8 lattice rod bundle to simulate scaled pressurized water reactor and boiling water reactor subchannel assemblies. To establish a high-spatial resolution database of the void fraction in the reactor fuel assembly geometries, tomographic measurements were performed with the High-Resolution Gamma-ray Tomography System, which was designed and built in house; the detector system has a spatial resolution of less than 1.0 mm using 240 LYSO (Lu1.8Y0.2SiO5) scintillators with a fan-beam array. In the present study, the local void fraction was measured with the MARBLE facility under various air-water flow conditions (jg = 0.04 to 0.85 m/s and jl = 0.12 to 0.77 m/s) covering from bubbly to cap-turbulent flows. The local void fraction was also successfully measured under nonuniform and asymmetric air bubble distribution conditions with an investigation of the effect of spacer grids and mixing vanes on void drift across subchannels.
