Sweng ng Woo

Comparative Study of Commercial CFD Software Performance for Prediction of Reactor Internal Flow

Even if some CFD software developers and its users think that a state-of-the-art CFD software can be used to reasonably solve at least single-phase nuclear reactor safety problems, there remain limitations and uncertainties in the calculation result. From a regulatory perspective, the Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of commercial CFD software for nuclear reactor safety problems. In this study, to examine the prediction performance of commercial CFD software with the porous model in the analysis of the scale-down APR (Advanced Power Reactor Plus) internal flow, a simulation was conducted with the on-board numerical models in ANSYS CFX R.14 and FLUENT R.14. It was concluded that depending on the CFD software, the internal flow distribution of the scale-down APR was locally somewhat different. Although there was a limitation in estimating the prediction performance of the commercial CFD software owing to the limited amount of measured data, CFX R.14 showed more reasonable prediction results in comparison with FLUENT R.14. Meanwhile, owing to the difference in discretization methodology, FLUENT R.14 required more computational memory than CFX R.14 for the same grid system. Therefore, the CFD software suitable to the available computational resource should be selected for massively parallel computations.

Numerical Analysis of Turbulent Flow around Tube Bundle by Applying CFD Best Practice Guideline

In this study, the numerical analysis of a turbulent flow around both a staggered and an inline tube bundle was conducted using ANSYS CFX V.13, a commercial CFD software. The flow was assumed to be steady, incompressible, and isothermal. According to the CFD Best Practice Guideline, the sensitivity study for grid size, accuracy of the discretization scheme for convection term, and turbulence model was conducted, and its result was compared with the experimental data to estimate the applicability of the CFD Best Practice Guideline. It was concluded that the CFD Best Practice Guideline did not always guarantee an improvement in the prediction performance of the commercial CFD software in the field of tube bundle flow.

Numerical Analysis of Internal Flow Distribution in Scale-Down APR+

A series of 1/5 scale-down reactor flow distribution tests had been conducted to determine the hydraulic characteristics of an APR+ (Advanced Power Reactor Plus), which were used as the input data for an open core thermal margin analysis code. In this study, to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ internal flow, simulations were conducted using the commercial multi-purpose computational fluid dynamics software ANSYS CFX V.14. It was concluded that the porous domain approach for some reactor internal structures could adequately predict the flow characteristics inside a reactor in a qualitative manner. If sufficient computational resources are available, the predicted core inlet flow distribution is expected to be more accurate by considering the real geometry of the internal structures, especially upstream of the core inlet.

Numerical Study on the Effect of Reactor Internal Structure Geometry Treatment Method on the Prediction Accuracy for Scale-down APR+ Flow Distribution

Abstract : Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core inlet flow rate distribution depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, to examine the effect of the reactor internal structure geometry treatment method on the prediction accuracy for the scale-down APR+ flow distribution, simulations with real geometry modeling were conducted using ANSYS CFX R.14, a commercial computational fluid dynamics software, and the predicted results were compared with those of the porous medium assumption. It was concluded that the core inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computational resources are available, an exact representation of these internal structures, for example, lower support structure bottom plate and ICI nozzle support plate, is needed for the accurate simulation of the reactor internal flow.

Sensitivity Study on Turbulence Models for the Prediction of the Reactor Internal Flow

In this study, in order to assess the prediction performance of Reynolds-averaged Navier-Stokes (RANS)-based turbulence models for the analysis of flow distribution inside the 1/5 scaled-down APR+ (Advanced Power Reactor Plus), the simulation was conducted with the commercial computational fluid dynamics software, ANSYS CFX R.13. The results predicted were then compared with the measured data. It was concluded that reactor internal-flow pattern differed locally; depending on the turbulence models used. In particular, the prediction performance of turbulence models showed the largest difference in the regions from the flow skirt to fuel assembly inlet. The prediction performance of the k-e model was superior to other turbulence models. This model also predicted the relatively uniform distribution of core-inlet flow-rate.Copyright

Numerical Analysis for the Effect of Flow Skirt Geometry on the Flow Distribution in the Scaledown APR

In this study, in order to examine the applicability of computational fluid dynamics with the porous model to the analysis of APR+ (Advanced Power Reactor Plus) internal flow, simulation was conducted with the commercial multi-purpose computational fluid dynamics software, ANSYS CFX V.14. In addition, among the various reactor internals, the effect of flow skirt geometry on reactor internal flow was investigated. It was concluded that the porous model for some reactor internal structures could adequately predict the hydraulic characteristics inside the reactor in a qualitative manner. If sufficient computation resource is available, the predicted core inlet flow distribution is expected to be more accurate, by considering the real geometry of the internal structures, especially located in the upstream of the core inlet. Finally, depending on the shape of the flow skirt, the flow distribution was somewhat different locally. The standard deviation of the mass flow rate () for the original shape of flow skirt was smaller, than that for the modified shape of flow skirt. This means that the original shape of the flow skirt may give a more uniform distribution of mass flow rate at the core inlet plane, which may be more desirable for the core cooling.

Evaluation of Potential to Air Ingestion From RWT Following RAS

At the Recirculation Actuation Signal (RAS) when the Refueling Water Tank (RWT) water level decreased to a certain value following Loss-of-Coolant Accident (LOCA), the isolation valves of Containment Recirculation Sump (CRS) of the Korean Standard Nuclear Power Plants (KSNP) are open automatically while the RWT isolation valves would be closed manually. It was concerned whether the design has a potential to air ingestion to Emergency Core Cooling System (ECCS) pumps before completion of the manual action to close RWT isolation valves. To support the safety evaluation on this issue including the evaluation of design adequacy, an analysis of the hydraulic transient within the ECCS piping following the RAS in KSNP is performed. RELAP5/MOD3.3 code is used to calculate the transient behavior of the piping network. The code was known to have capability to calculate one-dimensional two-phase transient flow with noncondensible gas in the complex piping. Substantial portion of ECCS are modeled including RWT, CRS, each pipe line from RWT and CRS to connection point with its own isolation valve and check valve, a common pipe line to ECCS header, each pipe line from the header to High Pressure Safety Injection (HPSI) pump, Low Pressure Safety Injection (LPSI) pump, and Containment Spray (CS) pump. Transient hydraulic behavior in the piping system following RAS after LOCA is calculated. It is found that the RWT water level was always higher than the elevation of the check valve at the connecting point by more than 15 ft. It indicates the air intrusion to the check valve can be sufficiently prevented by this amount of water head.