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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
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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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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.
S. Kliem, S. Danilin, A. Hämäläinen, J. Hádek, A. Keresztúri, P. Siltanen
Nuclear Science and Engineering | Volume 157 | Number 3 | November 2007 | Pages 280-298
Technical Paper | doi.org/10.13182/NSE07-A2728
Articles are hosted by Taylor and Francis Online.
Recently, three-dimensional neutron-kinetics core models have been coupled to advanced thermal-hydraulic system codes. These coupled codes can be used for the analysis of the whole reactor system. In the framework of the international association Atomic Energy Research (AER) on VVER Reactor Physics and Reactor Safety, two benchmarks for these code systems were defined. The reference reactor is the Russian VVER-440. The response of the reactor core to a symmetric and an asymmetric main steam line break should be investigated. So, different aspects of the coupling could be tested. As an additional feature, the participants had to use their own nuclear data.Each of these benchmarks was calculated by five different code systems. The comparison of the received solutions for the symmetric case shows good agreement in the evolution of the thermal hydraulics. When the core power reestablishes after recriticality, differences between the single solutions develop, mainly connected with the use of different nuclear data. Because of the increased complexity of the calculations, in the second benchmark differences between the thermal-hydraulic behavior in the single calculations were observed, additionally. These differences have their main origin in the behavior of the secondary side.The results of both benchmarks show the safety potential of the VVER-440 reactor. Even under very conservative conditions, no fuel rod failure was determined by the calculations, and the reactor was transferred into a subcritical final state.