Abstract
In the design basis accidents (DBA) of nuclear safety analysis, there are many thermal hydraulic phenomena must concern. When a loss-of-coolant accident (LOCA) occurs, the pressure inside the containment rises rapidly. The peak value of the pressure could undermine the containment integrity, causing containment shielding leakage and release of the radioactive material into the atmosphere. During a LOCA, the reactor core cladding zirconium reacts with water to produce a large amount of hydrogen gas, which is released into the containment. The accumulation of hydrogen gas concentrations may cause hydrogen combustion, even lead to the threat of a hydrogen explosion. This research focused on the common thermal hydraulic phenomena inside the reactor containment and analyzed the hydrogen diffusion behavior in the containment; and analyze whether hydrogen explosion or hydrogen combustion phenomenon. Another analyzed objects of this study the use of AP1000 advanced type light water pressure reactor and nuclear Tai-power Kuosheng, Maanshan nuclear power plants for the analysis of the subject. There are two issues in this study: 1) The GOTHIC code simulate ability of AP1000 passive system and, 2) The hydrogen diffusion model in containment. The first issue was focused on the containment safety of AP1000 advanced nuclear power plant. This study investigates the integrity of the protective mechanism of the AP1000 containment system during a LOCA was investigated. The performance of the passive containment cooling system (PCCS) and its ability to perform decay heat removal for long-term cooling were evaluated. The PCCS utilizes gravity-driven natural convection and naturally induced passive safety devices to release the decay heat of fuels into the atmosphere. In this study, two accidents were analyzed: a double-ended guillotine break in a hot leg and a double-ended break in a main steam line. The analytical results were compared with the corresponding results provided in the AP1000 ‘‘Design Control Document’’ (DCD). Short-term calculations and comparisons with the DCD suggested that GOTHIC 8.0 with a natural convection model may appropriately represent the phenomenon of passive, safe heat removal. In addition, the long-term calculations of spraying the outer primary containment with a water film and the steam condensation of the inner containment were also simulated. The second issue of the study also referred to containment thermal hydraulic safety analysis data for Taipower Maanshan nuclear power plant under a SBO accident. Based on the different containment structure data because of differences in the containment models, we can further analyze the state of the hydraulic behavior and hydrogen diffusion transfer mechanism. Based on the results, we can summarize a set of simple methodologies for the GOTHIC code, which can enhance the nuclear containment safety analysis ability in Taiwan. GOTHIC 8.0 code was used in the analysis and simulation, and the results, such as containment temperature pressure response, were verified through comparison with the corresponding results in DCD. The code also included Chen’s correlation, which gives the relationship of water film heat transfer. Chen’s model can calculate the heat transfer by liquid film evaporation convection at the secondary containment air flow path area.