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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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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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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.
Kwang Won Seul, Young Seok Bang, Hho Jung Kim
Nuclear Technology | Volume 126 | Number 3 | June 1999 | Pages 265-278
Technical Paper | Thermal Hydraulics | doi.org/10.13182/NT99-A2973
Articles are hosted by Taylor and Francis Online.
The potential of the RELAP5/MOD3.2 code was assessed for a loss-of-residual-heat-removal (RHR) event during midloop operation, and the predictability of major thermal-hydraulic phenomena was evaluated for the long-term transient. The calculations were compared for two cases of experiments conducted at the Rig of Safety Assessment-IV (ROSA-IV)/Large-Scale Test Facility (LSTF) in Japan: the cold-leg-opening and the pressurizer-manway-opening cases. In addition, the real plant responses to the event were evaluated for Yong Gwang nuclear power plant Units 3 and 4 (YGN 3/4) in Korea, especially concerning the mitigation capability to remove the decay heat through the steam generators (SGs). From the LSTF simulation, it was found that the RELAP5 code was capable of simulating the plant behavior following the loss-of-RHR event under a shutdown condition. As a result, the thermal-hydraulic transport process including noncondensable gas behavior was reasonably predicted with an appropriate time step and CPU time, and the major thermal-hydraulic phenomena agreed well with the experiment. However, there were some code deficiencies such as an estimation of large system mass errors for the long transient and severe flow oscillations in the core region. These should be improved for more accurate and reliable calculation. In the YGN 3/4 simulation, the water-filled SG case delayed the coolant discharge to containment by ~2 h and the core heatup by ~1.3 h, as compared to the emptied-SG case, because of reduction of the pressurization rate that resulted from condensation on the SG U-tube wall. For the water-filled SGs, the amount of heat transfer into the secondary side was estimated at more than 60% of the total core power throughout the transient.
