No abstract

Two-phase flow fields inside feeder pipes of a CANDU reactor have been simulated numerically using a CFD (computational fluid dynamics) code to calculate the shear stress distribution which is the most important factor to be considered in predicting the local areas of feeder pipes highly susceptible to FAC (flow-accelerated corrosion)-induced wall thinning. The CFD approach with schemes used in this study to simulate the turbulent flow situations inside the CANDU feeder pipes had been verified by showing a good agreement between the investigation results for the failed feedwater pipe at Surry Unit 2 plant in U.S. and the CFD calculation. Sensitivity studies of the three geometrical parameters such as angle of the 1st and 2nd bends, length of the 1st span between the grayloc hub and the 1st bend, and length of the 2nd span between the 1st and the 2nd bends had already been performed. In this study, the effects of void fraction of the primary coolant coming out from the exit of pressure tubes containing nuclear fuels on the fluid shear stress distribution at the inner surface of feeder pipe wall have been investigated to find out the local areas of feeder pipes conveying two-phase coolant, where are highly susceptible to FAC (flow-accelerated corrosion)-induced wall thinning. As the results of CFD analysis, it is seen that the local regions of feeder pipes of the operating CANDU reactors in Korea, on which the wall thickness measurements have been performed so far, are not coincided with the worst regions predicted by the present CFD analysis where is the connection region of straight & bend pipe near the inlet part of the bend intrados. Finally, based on the results of the present CFD analysis a guide to the selection of the weakest local positions where the measurement of wall thickness should be performed with higher priority has been provided.

When an incident of crude oil spill from an oil carrier occurs in the sea near the nuclear power plants, the spilled oil can be transported to the intake pit, where the NSCW (nuclear service cooling water) pumps locate, by sea current and wind drift (induced) current. The NSCW pumps take the essential service water from the sea being used as the ultimate heat sink and supply to the component cooling water heat exchangers to remove the decay heat generated from reactor. The NSCW system shall provide sufficient cooling capacity during the reactor normal operation, transients, and loss-of-coolant accidents (LOCAs). In this regard, it is very important to confirm if the cooling function of the NSCW system can be threaten due to spilled oil which may overspread the surface of sea water in the intake area. Thus, in this work, when an incident of crude oil spill occurs in the sea near a nuclear power plant using sea water as the ultimate heat sink, possibility of crude oil ingression into the component cooling water heat exchangers through the NSCW pumps has been evaluated in a conservative manner. To do this, for a flow field surrounding a NSCW pump equipped in an intake pit where a limited volume of sea water is initially contained and its surface is covered with spilled crude oil, a numerical simulation is performed using a CFD (Computational Fluid Dynamics) code. The objective of the present simulation is to find out the critical sea water level at or below which oil or air floating on the surface of sea water begins to be sucked in the pump inlet nozzle. The numerical simulation results show that in general, the oil covering the free surface of sea water in the intake can hardly flow into the pump inlet until the floating oil changes into tars which are heavier than oil as long as the sea water level does not fall below the limit value specified in the Technical Specification of each nuclear plant. It is also shown that the critical sea water level increases as the pump flowrate increases. This is physically plausible considering the effect of gravitational force.

Two-phase flow fields inside feeder pipes of a CANada Deuterium Uranium (CANDU) reactor have been simulated numerically using a computational fluid dynamics (CFD) code to calculate the shear stress distribution, which is the most important factor to be considered in predicting the local areas of feeder pipes highly susceptible to flow-accelerated corrosion (FAC)-induced wall thinning. The CFD approach with schemes used in this study to simulate the turbulent flow situations inside the CANDU feeder pipes has been verified by showing a good agreement between the investigation results for the failed feedwater pipe at Surry Unit 2 plant in the U.S. and the CFD calculation. Sensitivity studies of the three geometrical parameters such as angle of the first and second bends, length of the first span between the grayloc hub and the first bend, and length of the second span between the first and second bends had already been performed. In this study, the effects of void fraction of the primary coolant coming out from the exit of pressure tubes containing nuclear fuel on the fluid shear stress distribution at the inner surface of the feeder pipe wall have been investigated to find out the local areas of feeder pipes conveying a two-phase coolant, which are highly susceptible to FAC-induced wall thinning. From the results of the CFD analysis, it is seen that the local regions of feeder pipes of the operating CANDU reactors in Korea, on which the wall thickness measurements have been performed so far, do not coincide with the worst regions predicted by the present CFD analysis, which is the connection region of straight and bend pipes near the inlet part of the bend intrados. Finally, based on the results of the present CFD analysis, a guide to the selection of the weakest local positions where the measurement of wall thickness should be performed with higher priority has been provided.