Date of Award
Master of Science in Mechanical Engineering and Applied Mechanics
Mechanical, Industrial and Systems Engineering
Nuclear energy is the largest source of carbon-free electricity in the United States, making up 20 percent of the electricity generated in the United States. The United States is the second-largest energy-consuming country globally, with fossil fuels being the largest electricity-producing source. With Climate Change at the head of the world’s most difficult circumstances, it is evident that nuclear power is a crucial and significant source of carbon-free energy to combat this crisis. The NuScale Power SMR can provide a cost-effective and safe solution to further the expansion of nuclear energy throughout the United States and the world. The nature of the buoyancy-driven natural circulation cooling design of the reactor primary systems and the modularity and scalability power plant system provide the innovations and technology needed to do so. There are few tools like RELAP5-3D that allow for the thermal-hydraulic transient analysis of nuclear reactors. Due to the minimum amount of open literature available on the transient analysis of the NuScale Power SMR, RELAP5-3D has been utilized to perform the steady-state and a steam generator tube-rupture transient calculation. The benchmark experiment for thermal-hydraulic calculation codes, called Edward’s pipe blowdown experiment, was first modeled to understand the basics of a transient two-phase flow model. This experiment was performed to acquire the essential modeling skills and techniques to build the model and perform the calculations of the NuScale Power SMR using RELAP5-3D.
The NuScale Power Small Modular Reactor (SMR) relies on buoyancy-driven natural circulation cooling to cool the reactor core and extract thermal energy for electricity generation. The natural convection phenomenon has been of research interest for many years. NuScale Power LLC has only developed the SMR in recent years, and this integral Pressurized Water Reactor (iPWR) is the first nuclear reactor to utilize this phenomenon. Therefore, there is an increased interest in performing the transient analysis of the thermal-hydraulics of this reactor to understand conditions in which the natural circulation cooling inside the reactor system may be disrupted. There have been minimal published resources on this topic to date, making this research necessary for the growth and future of SMRs and natural circulation cooling of nuclear reactors. The innovations and designs of the NuScale Power SMR have allowed for enhanced safety, cost, scalability, modularity, time of construction, ease of transportation, and standardized manufacturing process of SMRs and nuclear power plants. RELAP5-3D was utilized to develop the model of the NuScale Power SMR and perform steady-state and transient analysis calculations of the reactor. This model was developed using the publicly available design data and parameters released by the U.S. NRC for the NuScale Power Final Safety Analysis Report (FSAR). The steady-state conditions of the reactor were modeled to simulate the reactor operation conditions in preparation for the transient analysis calculation. A tube rupture of the secondary steam generator was simulated for the transient analysis calculation to understand if the natural circulation cooling would be disrupted and if the secondary coolant would rise to dangerous levels proposing system failures.
The steady-state model simulated the proper reactor operational conditions, exhibiting higher mass flow rates than the best estimate flow rate specified in NuScale FSAR. The core temperatures were on the higher end of the temperature range but were still within the operational conditions, with the pressure controlled at 1850 psia. The forward flow energy loss coefficients proposed a particular issue in manipulating the code to obtain the core's correct mass flow rates and temperatures. It was found that the loss coefficients could be changed in a manner that lowered the mass flow rates closer to the best estimate flow rate, but the temperatures would, in turn, increase. Because the mass flow rate specified in NuScale FSAR was the best estimate value, the author concluded that the steady-state model was sufficient for the tube rupture model. The tube rupture was modeled using a single junction that connected the primary and secondary steam generators. The model was created to simulate a single helical coil steam generator tube being ruptured. Depressurization was not seen on the primary because the pressurizer was modeled as a pressure boundary condition at 1850 psia. A mass flow rate of approximately 36 lbm/s was seen through the tube rupture to the secondary side of the system. The water level did not increase significantly, but the liquid void fraction increased slightly. The flow through the rupture was choked because of the flashing of the liquid at high temperature and pressure to vapor at the lower pressure. It was found that instabilities and oscillations occurred very quickly on the primary and secondary sides, but the natural circulation flow was not disrupted.
Johnson, Kyle P., "TRANSIENT ANALYSIS OF THE NUSCALE POWER HELICAL-COIL STEAM GENERATOR TUBE RUPTURE USING RELAP5-3D" (2021). Open Access Master's Theses. Paper 1989.