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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
G7 pledges support for nuclear at Italy meeting
The Group of Seven (G7) recommitted its support for nuclear energy in the countries that opt to use it at a Ministerial Meeting on Climate in Italy last month.
In a statement following the April meeting, the group committed to support multilateral efforts to strengthen the resilience of nuclear supply chains, referencing the goal set by 25 countries during last year’s COP28 climate conference in Dubai to triple global nuclear generating capacity by 2050.
Hong Xu, Aurelian Florin Badea, Xu Cheng
Nuclear Technology | Volume 208 | Number 8 | August 2022 | Pages 1324-1336
Technical Paper | doi.org/10.1080/00295450.2021.2014755
Articles are hosted by Taylor and Francis Online.
The Primary Coolant Loop Test Facility [Primӓrkreislӓufe Versuchsanlage (PKL)] PKL I2.2 Benchmark experiment for an intermediate-break loss-of-coolant accident (IB-LOCA) with a 13% or 17% break at the cold leg was performed in the Organisation for Economic Co-operation and Development/PKL-4 project at PKL in Erlangen, Germany. Analysis of Thermal-Hydraulics of LEaks and Transients (ATHLET) 3.1A was used at Karlsruhe Institute of Technology for its posttest calculations. Crucial predicted parameters were compared with measured data. The calculated trend of the selected parameters fits well with that of the experimental data except for the phenomenon of core heatup and the value of the peak cladding temperature. A fast Fourier transform–based method was chosen to quantify the matching of the parameter trends. According to the quantitative assessment, the IB-LOCA scenario and its detailed phenomena can be predicted well by ATHLET. Additionally, some discrepancies, i.e., insufficient reliable predictions for break mass flow and for reactor pressure vessel collapsed water level, were also observed, possibly deserving another study to undergo deeper scrutiny.