Zhaofei Tian

Challenge Analysis and Schemes Design for the CFD Simulation of PWR

CFD simulation for a PWR is an important part for the development of Numerical Virtual Reactor (NVR) in Harbin Engineering University of China. CFD simulation can provide the detailed information of the flow and heat transfer process in a PWR. However, a large number of narrow flow channels with numerous complex structures (mixing vanes, dimples, springs, etc.) are located in a typical PWR. To obtain a better CFD simulation, the challenges created by these structural features were analyzed and some quantitative regularity and estimation were given in this paper. It was found that both computing resources and time are in great need for the CFD simulation of a whole reactor. These challenges have to be resolved, so two schemes were designed to assist/realize the reduction of the simulation burden on resources and time. One scheme is used to predict the combined efficiency of the simulation conditions (configuration of computing resources and application of simulation schemes), so it can assist the better choice/decision of the combination of the simulation conditions. The other scheme is based on the suitable simplification and modification, and it can directly reduce great computing burden.

Numerical Investigation of Single-Phase Heat Transfer in a 2 × 2 Rod Bundle with Spacer Grids for a High Pressure Heat Transfer Facility

Abstract Numerical simulation has been widely used in nuclear reactor safety analyses to gain insight into key phenomena. This paper compares simulations of a single-phase steady flow in a 2 × 2 rod bundle with spacer grids among different codes based on the high pressure heat transfer facility at University of Wisconsin. The detailed computational fluid dynamics modeling methodology was developed using FLUENT to help in the facility design and pretest analyses. After comparison between different turbulence models, the Standard k-ω was chosen to simulate the effect of unheated solid walls and grid spacers. It was found that solid walls had a small influence on the flow and heat transfer behavior. We note the effect of rod-to-wall gap needs be taken into account if it is larger than half of the gap between the rods. We compared the simulations of FLUENT, COBRA-TF, and TRACE to determine the position of thermocouples to be used in the planned experiments. An investigation was performed on the effect of bending angles of the grid spacer mixing vanes. Results showed that a larger bending angle results in higher turbulence mixing and locally higher Nusselt numbers downstream of the mixing vanes. Also, a small change of the bending angles results in a notable difference in the temperature distributions of the main flow.

Numerical investigation of the effect of grids and turbulence models on Critical Heat Flux in a vertical pipe

Numerical simulation has been widely used in nuclear reactor safety analyses to gain insight into key phenomena. The Critical Heat Flux (CHF) is one of the limiting criteria in the design and operation of nuclear reactors. It is a two-phase flow phenomenon, which rapidly decreases the heat transfer performance at the rod surface. This paper presents a numerical simulation of a steady state flow in a vertical pipe to predict the CHF phenomena. The detailed Computational Fluid Dynamic (CFD) modeling methodology was developed using FLUENT. Eulerian two-phase flow model is used to model the flow and heat transfer phenomena. In order to gain the peakwall temperature accurately and stably, the effect of different turbulence models and wall functions are investigated based on different grids. Results show that O type grid should be used for the simulation of CHF phenomenon. Grids with Y+ larger than 70 are recommended for the CHF simulation because of the acceptable results of all the turbulence models whileGrids with Y+ lower than 50 should be avoided.To predict the dry-out position accurately in a fine grid, Realizable k-e model with standard wall function is recommended.These conclusions have some reference significance to better predict the CHF phenomena of vertical pipe. It can also be expanded to rod bundle of Boiling Water Reactor (BWR) by using same pressure condition.

Simulation and Analysis of Small Break LOCA for AP1000 Using RELAP5-MV and Its Comparison with NOTRUMP Code

Many reactor safety simulation codes for nuclear power plants (NPPs) have been developed. However, it is very important to evaluate these codes by testing different accident scenarios in actual plant conditions. In reactor analysis, small break loss of coolant accident (SBLOCA) is an important safety issue. RELAP5-MV Visualized Modularization software is recognized as one of the best estimate transient simulation programs of light water reactors (LWR). RELAP5-MV has new options for improved modeling methods and interactive graphics display. Though the same models incorporated in RELAP5/MOD 4.0 are in RELAP5-MV, the significant difference of the latter is the interface for preparing the input deck. In this paper, RELAP5-MV is applied for the transient analysis of the primary system variation of thermal hydraulics parameters in primary loop under SBLOCA in AP1000 NPP. The upper limit of SBLOCA (10 inches) is simulated in the cold leg of the reactor and the calculations performed up to a transient time of 450,000.0 s. The results obtained from RELAP5-MV are in good agreement with those of NOTRUMP code obtained by Westinghouse when compared under the same conditions. It can be easily inferred that RELAP5-MV, in a similar manner to RELAP5/MOD4.0, is suitable for simulating a SBLOCA scenario.

Research on the Subchannel Analysis Method via CFD Analysis for PWR

Abstract In a subchannel analysis, the assumptions of the physical models may be invalid when three-dimensional (3-D) effects play an important role because a large-scale model cannot consider a small-scale physical process. However, in a pressurized water reactor (PWR), the flow process has a high 3-D effect due to the effect of complex structures, such as dimple, spring, and mixing vane. A computational fluid dynamics (CFD) analysis can give more detailed physical information. So, the modeling assumptions of the subchannel analysis codes were analyzed using data from CFD analysis, and some issues were found: The spatial acceleration of the cross-flow rate and the viscous force from fluid to fluid should not be neglected; the lateral pressure gradient not only is a driving force but also can be a resistance at some vertical range; the traditional “resistant force term” has the same direction with the cross flow at some vertical ranges. To improve the subchannel code, one physical term considering both the driving and the resistance effect is suggested to be added in the traditional transverse momentum equation. The solution for this new term and the method using spatial acceleration of the cross flow were also provided.