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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
2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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
Lightbridge announces first U-Zr fuel rod samples extruded at INL
Lightbridge Corporation announced today that it has reached “a critical milestone” in the development of its extruded solid fuel technology. Coupon samples using an alloy of zirconium and depleted uranium—not the high-assay low-enriched uranium (HALEU) that Lightbridge plans to use to manufacture its fuel for the commercial market—were extruded at Idaho National Laboratory’s Materials and Fuels Complex.
E. Uspuras, A. Kaliatka
Nuclear Technology | Volume 158 | Number 1 | April 2007 | Pages 18-25
Technical Paper | Best Estimate Methods | doi.org/10.13182/NT07-A3821
Articles are hosted by Taylor and Francis Online.
This paper evaluates the so-called weak heat conduction mechanism, i.e., the heat transfer from heated-up fuel channels in the radial direction to cooled channels through the adjacent graphite columns in the RBMK-1500 reactor. The influence of this mechanism on the calculation results for a long-term loss-of-coolant accident is investigated.Two possibilities for modeling the heat transfer in the radial direction, between adjacent graphite columns, using the system code RELAP5 are presented: (a) employing the interstructure heat conduction model built into the RELAP5-3D code and (b) employing the model of the reactor gas circuit, which supplies a mixture of gases into the reactor cavity.Both means allow one to predict the localized heatup in the RBMK core. However, the modeling of the reactor gas circuit in parallel with the reactor cooling circuit decreases the calculation time-step size quite significantly. The analysis results demonstrate the capability of the RELAP5-3D code to model heat conduction in the radial direction between different heat structures much more easily and a hundred times faster.