With the advent of more powerful, less expensive computing
resources, more and more attention is being given to Monte Carlo
techniques in design application. In many circles, stochastic
solutions are considered the next best thing to experimental data.
Statistical uncertainties in Monte Carlo calculations are typically
determined by the first...
Fission product transport in a Gen. IV Gas-Cooled Fast Reactor Plant utilizing vented fuel has been characterized using analytical and computational methods. The goal was to increase current understanding of fission product transport in helium-cooled GFRs using vented fuel and to provide a toolset for determining issues which may arise...
The Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system capable of removing decay heat directly from the reactor core. Its modularity makes it scalable for use in reactors with various power levels. Work has previously been completed to support inclusion of the DRACS in liquid metal reactors...
The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains detailed descriptions of many different reactor facilities. A large portion of these experiments have not been fully modeled due to the unavailability of computational power at the time of the experiment’s execution. With the advent of renewed interest in Sodium...
As human exploration of space continues, eventually sights will be set on establishing a permanent manned habitat on the surface of Mars. The success of such a base will depend on many new and developing technologies. Low mass and high energy density power production units will be of paramount importance....
The need for cheap reliable energy, while simultaneously avoiding uranium supply constraints makes uranium carbide (UC) fueled Gas Fast Reactors offer an attractive nuclear reactor design. In order to qualify the fuel, an enhanced understanding of the behavior of uranium carbide during operation is paramount. Due to a reduced re-solution...
A limited scope Probabilistic Risk Assessment (PRA) of a vented nuclear fuel system for a Generation IV Gas-Cooled Fast Reactor Plant was performed. The goal of the study was to better understand the safety and licensing implications of vented fuel technology. A Level 1 PRA was performed to determine the...