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Robotics & Remote Systems
The Mission of the Robotics and Remote Systems Division is to promote the development and application of immersive simulation, robotics, and remote systems for hazardous environments for the purpose of reducing hazardous exposure to individuals, reducing environmental hazards and reducing the cost of performing work.
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2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott 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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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
T. M. Conboy, T. J. McKrell, M. S. Kazimi
Nuclear Technology | Volume 182 | Number 3 | June 2013 | Pages 259-273
Technical Paper | Fuel Cycle and Management/Thermal Hydraulics | doi.org/10.13182/NT12-58
Articles are hosted by Taylor and Francis Online.
In order to significantly increase the power density of light water reactors (LWRs), the authors propose the helical-cruciform (HC) fuel rod assembly as an alternative to traditional fuel geometry. The HC fuel rod assembly is a self-supporting nuclear fuel configuration consisting of four-petalled, axially twisted fuel rods closely packed against one another in a square array. Within the LWR core, HC fuel would possess several advantages over traditional fuel, potentially allowing for operation at a higher power density. Chief among these advantages are a larger surface-to-volume ratio, improved radial mixing characteristics of the coolant, and a shorter radial heat conduction path in the fuel pellet. In adapting helical rod geometry to the LWR core, the authors identified a shortage of correlations for fluid flow in twisted geometry flow channels, causing uncertainty in modeling studies. This gap was addressed by constructing an experimental facility for the measurement of hydraulic resistance and assembly mixing within a mock bundle of HC fuel rods. The rods were manufactured and tested in 4 × 4 square arrays at twist pitches of 200, 100, and 50 cm. Hydraulic resistance was evaluated by measuring frictional pressure drop over a 1-m length of the assembly. Results showed a higher pressure drop for the HC rods in comparison to bare cylindrical rods with no spacers, at a given mass flux, but no apparent dependency on twist pitch. However, data indicated that the HC-rod effective hydraulic diameter was only 90% of the expected value given its wetted perimeter and flow area, suggesting a shift from the traditional definition of Dh for this unique shape. Mixing tests used the technique of a hot water tracer injection into the central subchannel of the assembly of room-temperature water. Downstream temperature measurements were used to judge the rate of lateral cross flow within the HC rod bundle. Over 300 tests were analyzed, yielding a best-fit correlation for use with any twist pitch, rod length, or coolant mass flux. Compared to a traditional rod bundle, this correlation implies an enhancement in the intensity of turbulent interchange of 40% brought about by the HC geometry and a 1.6% forced diversion of axial flow per subchannel, per quarter-turn along the rod length. The correlations for hydraulics and cross-flow mixing presented here should reduce the uncertainty in future analysis of this fuel type for high-power-density LWRs.
