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Division Spotlight
Operations & Power
Members focus on the dissemination of knowledge and information in the area of power reactors with particular application to the production of electric power and process heat. The division sponsors meetings on the coverage of applied nuclear science and engineering as related to power plants, non-power reactors, and other nuclear facilities. It encourages and assists with the dissemination of knowledge pertinent to the safe and efficient operation of nuclear facilities through professional staff development, information exchange, and supporting the generation of viable solutions to current issues.
Meeting Spotlight
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C 2025)
April 27–30, 2025
Denver, CO|The Westin Denver Downtown
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
TerraPower begins U.K. regulatory approval process
Seattle-based TerraPower signaled its interest this week in building its Natrium small modular reactor in the United Kingdom, the company announced.
TerraPower sent a letter to the U.K.’s Department for Energy Security and Net Zero, formally establishing its intention to enter the U.K. generic design assessment (GDA) process. This is TerraPower’s first step in deployment of its Natrium technology—a 345-MW sodium fast reactor coupled with a molten salt energy storage unit—on the international stage.
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.
