ANS is committed to advancing, fostering, and promoting the development and application of nuclear sciences and technologies to benefit society.
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Division Spotlight
Reactor Physics
The division's objectives are to promote the advancement of knowledge and understanding of the fundamental physical phenomena characterizing nuclear reactors and other nuclear systems. The division encourages research and disseminates information through meetings and publications. Areas of technical interest include nuclear data, particle interactions and transport, reactor and nuclear systems analysis, methods, design, validation and operating experience and standards. The Wigner Award heads the awards program.
Meeting Spotlight
Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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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February 2025
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Latest News
Article considers incorporation of AI into nuclear power plant operations
The potential application of artificial intelligence to the operation of nuclear power plants is explored in an article published in late December in the Washington Examiner. The article, written by energy and environment reporter Callie Patteson, presents the views of a number of experts, including Yavuz Arik, a strategic energy consultant.
Yassin A. Hassan, Sibashis S. Banerjee
Nuclear Technology | Volume 108 | Number 2 | November 1994 | Pages 191-206
Technical Paper | Nuclear Reactor Safety | doi.org/10.13182/NT94-A35030
Articles are hosted by Taylor and Francis Online.
A simulation of the loss of residual heat removal (RHR) system during midloop operations was performed using the RELAP5/MOD3 thermal-hydraulic code. The experiment was conducted at the Rig of Safety Assessment (ROSA)-IV/Large-Scale Test Facility. The experiment involved a 5% cold-leg break along with the loss of the RHR system. The transient was simulated for 3040 s. Core boiling and subsequent primary system pressurization occurred after the initiation of the transient. There was a good agreement between the measured and the calculated data until the loop seal clearing (LSC). It was found that the steam condensation was underpredicted in the calculations. This caused the calculated data after the LSC to differ from that of the measured data. The core rod surface temperature excursion around the occurrence of the LSC was not calculated. Overall, there was good qualitative agreement between the measured and the calculated data. The calculations, performed on the CRAY-YMP supercomputer, took over 60 h of CPU time for a transient of 51 min.