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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.
Toshiharu Muramatsu, Hisashi Ninokata
Nuclear Technology | Volume 113 | Number 1 | January 1996 | Pages 54-72
Technical Paper | Heat Transfer and Fluid Flow | doi.org/10.13182/NT96-A35199
Articles are hosted by Taylor and Francis Online.
Two thermohydraulics computer programs AQUA and DINUS-3, which are represented by both time- and volume-averaged transport analysis and direct numerical simulation of turbulence, respectively, were developed and validated for the evaluation of thermal striping phenomena. These codes were incorporated with higher order difference schemes to approximate the convection terms in conservation equations and adaptive time-step size control systems based on the fuzzy theory to eliminate numerical instabilities. From validation analyses with fundamental experiments in water and sodium, it was concluded that (a) thermal striping conditions such as spatial distributions of the intensity and the frequency of the fluid temperature fluctuations can be estimated efficiently by a combined approach incorporating the AQUA code and the DINUS-3 code, and (b) the thermal striping phenomena for the in-vessel components of actual liquid-metal-cooled fast reactors can be evaluated by the numerical method without conventional approaches such as large scale model experiments using sodium.