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The Education, Training & Workforce Development Division provides communication among the academic, industrial, and governmental communities through the exchange of views and information on matters related to education, training and workforce development in nuclear and radiological science, engineering, and technology. Industry leaders, education and training professionals, and interested students work together through Society-sponsored meetings and publications, to enrich their professional development, to educate the general public, and to advance nuclear and radiological science and engineering.
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2024 ANS Annual Conference
June 16–19, 2024
Las Vegas, NV|Mandalay Bay Resort and Casino
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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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Remembering Joseph M. Hendrie
Joseph M. Hendrie
To those of us who knew Joe, even prior to his appointment as chair of the Nuclear Regulatory Commission, it is an understatement to say that he was a larger-than-life member of the nuclear science and technology enterprise. He was best known to the broader community for two major accomplishments: the design and construction of the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory and the creation of the standard review plan (SRP) for the U.S. Atomic Energy Commission.
In addition to the products of these endeavors becoming major fundaments to their respective communities, they were uniquely Joe. The safety analysis report for the HFBR was written essentially single-handedly by him. This was true of the SRP as well, which became the key safety review document for the NRC as it performed safety reviews for the growing number of power reactor applications in the United States. His deep technical knowledge of nuclear engineering and his extraordinary management skills made this possible.
Ki Yong Choi, Hyun Sik Park, Seok Cho, Kyoung Ho Kang, Nam Hyun Choi, Won Pil Baek, Yeon Sik Kim
Nuclear Technology | Volume 175 | Number 3 | September 2011 | Pages 604-618
Technical Paper | NURETH-13 Special / Thermal Hydraulics | doi.org/10.13182/NT11-A12509
Articles are hosted by Taylor and Francis Online.
The direct vessel injection (DVI)-adopted power plant APR1400 considers a DVI line break among the analyzed small-break loss-of-coolant accidents in safety analysis. The first-ever integral effects test database for various DVI line break sizes from 5% to 100% was established with the Korea Atomic Energy Research Institute's Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) test facility. This database enhances our physical understanding of the major thermal-hydraulic behaviors of the APR1400 during DVI line break accidents, and it can also be used to examine the prediction capabilities and identify any deficiencies in the existing best-estimate safety analysis codes. Effects of the break size were experimentally investigated, and the best-estimated MARS code was assessed against the experimental database. On the whole, the prediction of the MARS code shows a good agreement with the measured data. However, the code predicted a higher core level than the data just before a loop seal clearing occurs, and it also produced a more rapid decrease in the downcomer water level than the data. These disagreements are the expected consequence of uncertainties in predicting countercurrent flow or condensation phenomena in a downcomer region. The present integral effects test data will be used to support the present conservative safety analysis methodology and to develop a new best-estimate safety analysis methodology on the DVI line break accidents of the APR1400.