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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
2025 ANS Annual Conference
June 15–18, 2025
Chicago, IL|Chicago Marriott Downtown
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
Smarter waste strategies: Helping deliver on the promise of advanced nuclear
At COP28, held in Dubai in 2023, a clear consensus emerged: Nuclear energy must be a cornerstone of the global clean energy transition. With electricity demand projected to soar as we decarbonize not just power but also industry, transport, and heat, the case for new nuclear is compelling. More than 20 countries committed to tripling global nuclear capacity by 2050. In the United States alone, the Department of Energy forecasts that the country’s current nuclear capacity could more than triple, adding 200 GW of new nuclear to the existing 95 GW by mid-century.
Michitsugu Mori
Nuclear Technology | Volume 128 | Number 2 | November 1999 | Pages 205-215
Technical Paper | RETRAN | doi.org/10.13182/NT99-A3025
Articles are hosted by Taylor and Francis Online.
The advanced boiling water reactor (ABWR) has ten reactor-internal pumps peripherally mounted on the bottom of a reactor vessel. Analytical simulation of reactor-internal pumps unique to the ABWR requires new modeling because of the difference in core flow characteristics between the reactor-internal pumps and the two external-recirculation pumps of the primary outer loops with the jet pumps in a current boiling water reactor. Efforts in this work focused on modeling and simulation of reactor-internal pumps and core flow of the ABWR, using the RETRAN-3D code, the computer program for transient thermal-hydraulic analysis of a complex fluid flow system, without multidimensional kinetics. Included are modeling of the core and reactor pressure vessel with ten reactor-internal pumps, and simulation of the events of reactor-internal-pumps trip during the startup-phase tests, which are unable to be done in the simulation of a current BWR. Sensitivity analyses on the recirculation flow control and the slip model were also performed. The predictions by the RETRAN-3D code successfully tracked the measured data of reactor-internal-pump trip during the startup-phase test. The present analytical simulations could demonstrate the validation of the RETRAN-3D code applicable to the ABWR with the pump model of reactor-internal pumps in the program.