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Division Spotlight
Radiation Protection & Shielding
The Radiation Protection and Shielding Division is developing and promoting radiation protection and shielding aspects of nuclear science and technology — including interaction of nuclear radiation with materials and biological systems, instruments and techniques for the measurement of nuclear radiation fields, and radiation shield design and evaluation.
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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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.
Sandra M. Sloan, Yassin Hassan
Nuclear Technology | Volume 100 | Number 1 | October 1992 | Pages 111-124
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT92-A34757
Articles are hosted by Taylor and Francis Online.
The thermal-hydraulics simulation codes RELAP5/MOD2 and RELAP5/MOD3 are utilized to calculate the phenomena that occurred during a small-break loss-of-coolant accident (LOCA) simulation conducted at the ROSA-IV Large-Scale Test Facility. The objectives of the work are to analyze RELAP5/MOD2 and RELAP5/MOD3 predictions of a small-break LOCA simulation and to compare the ability of each code version to accurately predict the important physical phenomena of the experiment. The RELAP5/MOD2 and RELAP5/MOD3 predictions are compared with each other and assessed against the experimental results. The overall conclusion is that both code versions predict trends well, but each differs in the prediction of the magnitude and timing of occurrences. Specific areas of difference include primary system pressure, differential pressure in the upper plenum, core liquid level depression and subsequent heatup, core void fraction profile, and the differential pressure in the steam generator inlet plenum. All but the last of these differences are related to the RELAP5/MOD3 prediction of excessive liquid holdup in the upper plenum during the first core liquid depression, which is believed to lead to the prediction of water trickling into the upper core volumes and providing a cooling mechanism not present during the experiment. The liquid holdup is believed to be the result of an overprediction of interphase drag at the junctions between the upper plenum volumes. The trends of increase and decrease in steam generator liquid inventory are more correctly calculated by RELAP5/MOD3 than RELAP5/MOD2 because of the implementation of a countercurrent flow limitation equation at the inlet to the steam generator U-tubes. Although the results of any single exercise are not sufficient to make a global assessment of code performance capabilities, this study identifies an area that should be investigated more fully in future code assessment exercises by utilizing experimental data from transients that will further exercise the interphase drag computation capability of the code.
