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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
College students help develop waste-measuring device at Hanford
A partnership between Washington River Protection Solutions (WRPS) and Washington State University has resulted in the development of a device to measure radioactive and chemical tank waste at the Hanford Site. WRPS is the contractor at Hanford for the Department of Energy’s Office of Environmental Management.
Zheng Fu, Joshua Pack, Fatih Aydogan
Nuclear Science and Engineering | Volume 182 | Number 1 | January 2016 | Pages 119-134
Technical Paper | Special Issue on the RELAP5-3D Computer Code | doi.org/10.13182/NSE15-4
Articles are hosted by Taylor and Francis Online.
In the study and design of a nuclear power plant, extensive system modeling is necessary to determine how the reactor will perform in any given situation, not only in the normal performance of the reactor, but also in transients including unanticipated transients without scram and hypothetical accidents. One type of nuclear power plant under study is the hybrid energy system, which uses nuclear power to generate both electricity and heat for facilities. Obviously, the second steam cycle in the nuclear power plant requires several design updates and experiments. Unfortunately, the current versions of the Reactor Excursion and Leak Analysis Program (RELAP) do not allow online data streams from experimental facilities to the computational model of the secondary steam loop. Therefore, this study develops a coupling between RELAP5 and Laboratory Virtual Instrument Engineering Workbench (LabVIEW) to model primary and secondary coolant loops. In this way, the LabVIEW model can easily be connected to an experimental apparatus to provide an online data stream and the online transient behavior of an entire nuclear power plant system. This study shows two different coupling approaches and makes qualitative and quantitative comparisons between these approaches.
This paper demonstrates the results of different couplings between the primary and secondary systems of a typical pressurized water reactor (PWR). The primary loop model is a four-loop PWR. The model has been executed with steady state and transients (in this case, a loss-of-coolant accident). The results of both coupling methods have been compared with the typical RELAP5 results.