Electronic Thesis and Dissertation Repository

Thesis Format

Integrated Article

Degree

Doctor of Philosophy

Program

Mechanical and Materials Engineering

Supervisor

Zhang, Chao

Abstract

Canada participated in the Generation IV nuclear reactors with the Supercritical Water-Cooled Reactor (SCWR) concept. This work focuses on the numerical studies of the fluid flow and heat transfer of the supercritical water in the nuclear reactor fuel bundle, and the construction of the linear dynamic model and the design of the control system for the Canadian SCWR power plant.

Firstly, the fluid flow and heat transfer of the supercritical water in the vertical tube and the rod bundle is numerically investigated to evaluate whether the existing turbulent models could successfully caption the wall temperature variations at supercritical conditions by comparing the numerical results with the experimental data. The turbulent models that have better performance are modified using a variable turbulent Prandtl number model. The application of the proposed turbulence model shows a great improvement in the prediction of the wall temperatures under supercritical conditions. Accordingly, the full-scale simulation of the fluid flow and heat transfer of the supercritical water flow in the reactor fuel rod bundle was performed by the proposed turbulent model. The results show that the circumferential cladding surface temperature distribution is extremely non-uniform and the maximum cladding surface temperature for each fuel rod also shows large differences. In addition, the effects of operating conditions on the heat transfer of upward supercritical water flow in the reactor fuel bundle are studied numerically. The wall temperatures generally increase with the increase in the inlet temperature, heat flux, or the decrease in the mass flux. Buoyancy-affected zones mainly exist at the region around the pseudocritical temperature.

In this work, the design of the feedback control system for the SCWR is also carried out. The dynamic relationships between inputs and outputs of the reactor are obtained through transient computational fluid dynamics (CFD) simulations. The designed feedback control system can regulate the reactor back to the design point timely. Finally, a linear dynamic model for the entire SCWR power plant is developed, which includes the reactor, feedwater pump, outlet plenum, main steam line, turbine, and condenser. The dynamic characteristics of the system and the steady-state interaction between different inputs and outputs of the system are analyzed. The control system for the SCWR power plant is constructed and the performance of the control system is satisfactory.

Summary for Lay Audience

As the growth of earth’s population and the adverse impacts from global climate change, the demand for energy is increasing. Nuclear energy is prominent among all energy supplies with the advantages of clean, safe and cost-effective under appropriate use. The Supercritical Water-Cooled Reactor (SCWR) concept is one of six Generation IV nuclear reactors, which uses supercritical water as the coolant. In the reactor core, the heat produced by the nuclear fission process is absorbed by coolant. Given the peculiarity of thermophysical properties of supercritical water and the geometry of the flow channel, the heat transfer phenomenon in the supercritical water rod bundle is still not well-understood.

Firstly, the heat transfer of supercritical water in different vertical channels is numerically investigated by the existing turbulent models. Considering operating conditions and variations of thermophysical properties, a new variable turbulent Prandtl number model is developed to describe the heat transfer characteristics. Next, the numerical simulation of the upward supercritical water flow in the reactor rod bundle is performed to investigate the fluid flow and variations of wall temperatures. The gradient of the cladding surface temperature along the circumference is found large. Furthermore, influences of operating pressure, inlet temperature, heat flux, and mass flux on the heat transfer in the reactor fuel bundle are studied. The findings of this study show that wall temperature generally increases as the increase in the inlet temperature, heat flux, or the decrease in the mass flux. On the other hand, buoyancy-affected zones mainly exist at the region where thermophysical properties of supercritical water exhibit sharp changes. In addition, the buoyancy-affected zone is reduced with the increase in the inlet temperature, heat flux, and mass flux. This work also presents the design of a feedback control system for the SCWR. It is found that the designed control system can regulate the reactor to the original operating point timely when the reactor is subjected to disturbances. Finally, the control system for the entire SCWR power plant is also constructed. The performance of the control system is evaluated and it is satisfactory.

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