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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
ANS Student Conference 2025
April 3–5, 2025
Albuquerque, NM|The University of New Mexico
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
Norway’s Halden reactor takes first step toward decommissioning
The government of Norway has granted the transfer of the Halden research reactor from the Institute for Energy Technology (IFE) to the state agency Norwegian Nuclear Decommissioning (NND). The 25-MWt Halden boiling water reactor operated from 1958 to 2018 and was used in the research of nuclear fuel, reactor internals, plant procedures and monitoring, and human factors.
Ji-Hwan Hwang (Chung-Ang Univ), Min Ho Lee, In Cheol Bang (UNIST), Dong-Wook Jerng (Chung-Ang Univ)
Proceedings | 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) | Charlotte, NC, April 8-11, 2018 | Pages 650-656
A concept of the ERVC (External Reactor Vessel Cooling) can be applied to the Sodium Fast Reactor (SFR), by using air instead of water as a coolant. The RVCS, which is a system to maintain integrity of concrete structures, can be used for reactor exterior surface cooling. The heat removal by RVCS operation affects and affected by the natural circulation inside sodium pool. Thus, understanding the natural circulation inside sodium pool is important for RVCS performance prediction. In this paper, we numerically investigated similarity laws to figure out the applicability of water tests to actual sodium condition using a commercial CFD code, STAR-CCM+. In this study, 4 different scales, 1/20, 1/10, 1/8 and 1/5 were investigated. In every case the volumetric heat flux of core was identical. For numerical simulation, the geometry and configuration of Prototype Gen-IV Sodium Cooled Fast Reactor (PGFSR) developed in KAERI is chosen as reference, and modeled. For comparison, each similarity laws were compared in terms of temperature field. The simulation and comparison results show that by preserving modified Bousinessq number and Peclet number when reducing the test, the temperature field of reactor can be reproduced. However, the test is too small, the flow resistance due to internal structure acts dominantly, disturbing the flow. Therefore, for better result, such flow resistance which can be occurs at small-scaled experiment, should be taken into account.