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The human factor in licensing and operating the next generation of nuclear plants
As human factors specialists working at the intersection of human performance and nuclear operations, we are witnessing one of the nuclear sector’s most significant transitions in decades. The emergence of small modular reactors, microreactors, and other advanced designs is reshaping the industry’s landscape. Digital instrumentation and controls, passive safety systems, and increased automation are creating opportunities for greater safety margins and more flexible operation. These same features also fundamentally redefine what it means to “operate” a nuclear plant. Interactions among human roles, automation, and passive systems shape how people maintain awareness, exercise judgment, and intervene when necessary. These developments affect both operational realities and the regulatory foundations on which nuclear safety is built.
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.
