Xuewu Cao

Study on intermittent flow behavior in a vertical channel under low-pressure condition

The intermittent flow behavior in a vertical annulus under a low-pressure condition was experimentally studied using a scaling experiment facility. The temperature and pressure variations in the channel had been obtained under the heat load ranging from 0 to 2.0 kW, initial subcooled water temperature ranging from 50 to 90 °C and length–diameter ratio ranging from 1.6 to 50. The effects of the heat load and length–diameter ratio of channel on the flow characteristics were investigated in detail. The experimental results showed that the steam bubbles erupted more frequently and regularly at a high heat load. The intermittent flow period decreased with increase of the heat load and aspect ratio. Based on the mechanism analysis, an empirical model considering the steam oscillation and the vapor–liquid interface rupture based on the experimental data was proposed. It was found that the accumulated steam basically increased linearly. The oscillation of the pressure and velocity decreased gradually with continuous steam accumulation. The Reynolds number of the liquid within the rising section was very small at the stagnation state since there was no forced circulation flow. Finally, a blockage was engendered in the pipeline with the steam accumulated.

Study of Decay Heat Removal Approach for Advanced Passive PWR During Station Blackout

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.© 2014 ASME

Flow Transient in Primary Coolant System During Reactor Coolant Pump Start-Up Period

The reactor coolant pump is an integral part of the reactor coolant pressure boundary. Its security rank is the highest level. The flow transient analysis is very necessary in the reactor coolant pump design and the nuclear reactor design. Based on the momentum conservation and momentum balance relations, The transient flow rate and the pump speed during a pump start-up are derived The analytical flow rate, the pump speed and the kinetic energy stored in the rotating parts are all non-dimensionalized. A comparison with Tsukamoto’s experimental results during the pump start-up shows good agreement. The curves of non-dimensional start-up flow rate, the pump speed and the kinetic energy stored in the rotating parts for different system parameter β are predicted and compared. The effect of β on the flow rate, the pump speed and the kinetic energy stored in the rotating parts is discussed according to the inertia of primary loop fluid and the pump moment of inertia. In addition, the prediction of the flow rate, the pump speed transient of Qinshan and Dayawan reactor coolant pumps and Takada’s test pump during start-up period are also performed.© 2013 ASME

Preliminary Analysis of Effect of the Intentional Depressurization on Fission Product Behavior During TMLB’ Severe Accident

It has been identified that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH). Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its effects. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB’ base case and opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs, but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomena of revaporization is strong in the RCS.Copyright