Young Seok Bang

Analysis of experiments for in-tube steam condensation in the presence of noncondensable gases at a low pressure using the RELAP5/MOD3.2 code modified with a non-iterative condensation model

Abstract The standard RELAP5/MOD3.2 code is modified using non-iterative modeling, which is a mechanistic model developed for easy engineering application to simulate steam condensation in the presence of noncondensable gases in a tube. To predict the liquid-side heat transfer coefficients in the modified RELAP5/MOD3.2 code, Nusselt’s correlation is used for the condensation in a vertical tube and Kim’s correlation, correlated with the Froude number is used for the condensation in a horizontal tube. In the modified code, the wall friction in a vertical tube is calculated using the two-phase friction factor correlation proposed by Collier and the interfacial friction factor is calculated using the empirical power-law relationship of Choi. Both the standard and the modified RELAP5/MOD3.2 codes are used to simulate two kinds of vertical in-tube experiments and a horizontally stratified in-tube experiment involving the condensation phenomenon in the presence of noncondensable gases. Two vertical in-tube experiments, Passive Containment Cooling System (PCCS) condensation and reflux condensation experiments, provide data on the steady-state behaviors with a typical flow, pressure and air mass fraction conditions likely to be seen in a condensing tube of PCCS and in a U-tube of a steam generator in a mid-loop operation. The horizontally stratified in-tube experiment represents the direct-contact condensation phenomena in a hot leg of a nuclear reactor. The modeling capabilities of the modified code as well as the standard code for steam condensation in the presence of noncondensable gases are assessed using these three KAIST condensation experiments. The modified code gives a better prediction for the data of the three condensation experiments than that of the standard code. Simulation results of PCCS and reflux condensation experiments show that the local heat transfer coefficients are predicted well with the modified code but in the standard version of the code they are under-predicted by the default model and over-predicted by the alternative model. The modified code predicts well the experimental void fraction using the two-phase wall friction factor correlation. Simulation results of the horizontally stratified condensation experiments show that the modified code predicts the interfacial heat transfer coefficient better than that of the standard code.

A PARTICLE TRACKING MODEL TO PREDICT THE DEBRIS TRANSPORT ON THE CONTAINMENT FLOOR

An analysis model on debris transport in the containment floor of pressurized water reactors is developed in which the flow field is calculated by Eulerian conservation equations of mass and momentum and the debris particles are traced by Lagrange equations of motion using the pre-determined flow field data. For the flow field calculation, two-dimensional Shallow Water Equations derived from Navier Stokes equations are solved using the Finite Volume Method, and the Harten-Lax-van Leer scheme is used for accuracy to capture the dry-to-wet interface. For the debris tracing, a simplified twodimensional Lagrangian particle tracking model including drag force is developed. Advanced schemes to find the positions of particles over the containment floor and to determine the position of particles reflected from the solid wall are implemented. The present model is applied to calculate the transport fraction to the Hold-up Volume Tank in Advanced Power Reactors 1400. By the present model, the debris transport fraction is predicted, and the effect of particle density and particle size on transport is investigated.

Plant Behavior Following a Loss-of-Residual-Heat-Removal Event Under a Shutdown Condition

The potential of the RELAP5/MOD3.2 code was assessed for a loss-of-residual-heat-removal (RHR) event during midloop operation, and the predictability of major thermal-hydraulic phenomena was evaluated for the long-term transient. The calculations were compared for two cases of experiments conducted at the Rig of Safety Assessment-IV (ROSA-IV)/Large-Scale Test Facility (LSTF) in Japan: the cold-leg-opening and the pressurizer-manway-opening cases. In addition, the real plant responses to the event were evaluated for Yong Gwang nuclear power plant Units 3 and 4 (YGN 3/4) in Korea, especially concerning the mitigation capability to remove the decay heat through the steam generators (SGs). From the LSTF simulation, it was found that the RELAP5 code was capable of simulating the plant behavior following the loss-of-RHR event under a shutdown condition. As a result, the thermal-hydraulic transport process including noncondensable gas behavior was reasonably predicted with an appropriate time step and CPU time, and the major thermal-hydraulic phenomena agreed well with the experiment. However, there were some code deficiencies such as an estimation of large system mass errors for the long transient and severe flow oscillations in the core region. These should be improved for more accurate and reliable calculation. In the YGN 3/4 simulation, the water-filled SG case delayed the coolant discharge to containment by ∼2 h and the core heatup by ∼1.3 h, as compared to the emptied-SG case, because of reduction of the pressurization rate that resulted from condensation on the SG U-tube wall. For the water-filled SGs, the amount ofheat transfer into the secondary side was estimated at more than 60% of the total core power throughout the transient.

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.

Mitigation Measures Following a Loss-of-Residual-Heat-Removal Event During Shutdown

The transient following a loss-of-residual-heat-removal event during shutdown was analyzed to determine the containment closure time (CCT) to prevent uncontrolled release of fission products and the gravity-injection path and rate (GIPR) for effective core cooling using the RELAP5/MOD3.2 code. The plant conditions of Yonggwang Units 3 and 4, a pressurized water reactor (PWR) of 2815-MW(thermal) power in Korea, were reviewed, and possible event sequences were identified. From the CCT analysis for the five cases of typical plant configurations, it was estimated for the earliest CCT to be 40 min after the event in a case with a large cold-leg opening and emptied steam generators (SGs). However, the case with water-filled SGs significantly delayed the CCT through the heat removal to the secondary side. From the GIPR analysis for the six possible gravity-injection paths from the refueling water storage tank (RWST), the case with the injection point and opening on the other leg side was estimated to be the most suitable path to avoid core boiling. In addition, from the sensitivity study, it was evaluated for the plant to be capable of providing the core cooling for the long-term transient if nominal RWST water is available. As a result, these analysis methods and results will provide useful information in understanding the plant behavior and preparing the mitigation measures after the event, especially for Combustion Engineering-type PWR plants. However, to directly apply the analysis results to the emergency procedure for such an event, additional case studies are needed for a wide range of operating conditions such as reactor coolant inventory, RWST water temperature, and core decay heat rate.

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.