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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.
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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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Proving DRACO will deliver
The United States is now closer than it has been in over five decades to launching the first nuclear thermal rocket into space, thanks to DRACO—the Demonstration Rocket for Agile Cislunar Orbit.
Yoshihisa Nishi, Nobuyuki Ueda, Izumi Kinoshita, Ehud Greenspan
Nuclear Technology | Volume 152 | Number 3 | December 2005 | Pages 324-338
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT05-A3680
Articles are hosted by Taylor and Francis Online.
The encapsulated nuclear heat source (ENHS) is a modular reactor that was selected by the 1999 U.S. Department of Energy Nuclear Energy Research Initiative program as a candidate Generation IV reactor concept. It is a fast neutron spectrum reactor cooled by lead-bismuth eutectic using natural circulation. One of the unique features of the ENHS is that the fission-generated heat is transferred from the primary coolant to the secondary coolant through rectangular intermediate heat exchanger (IHX) channels. The decay heat is removed by the reactor vessel auxiliary cooling system (RVACS).Events of protected loss of heat sink (PLOHS) and unprotected transient overpower (UTOP) have been analyzed for the ENHS using the CERES transient simulation code for liquid-metal-cooled reactors.It is found that the ENHS core is sufficiently cooled by the RVACS under the PLOHS condition. The core flow rate is affected by the growth and disappearance of temperature stratification in the primary plenum. It is also found that even under the inconceivable UTOP event considered, the ENHS reactor core is not catastrophically damaged. This is due to negative reactivity feedback from the radial expansion of the core, the grid plate, and the Doppler effect. The use of high-performance ferritic steel instead of HT-9 and proper design of the reactor control system could provide large safety margins against cladding damage.