Jiyang Yu

PCCSAC: A Three-Dimensional Code for AC600 Passive Containment Cooling System Analysis

Abstract The algorithm for the transient containment analysis code (PCCSAC) developed to analyze the AC600 passive containment cooling system (PCCS) is described. The AC600 reactor uses passive external spray from an elevated tank. The PCCSAC code has been developed with a number of unique modeling capabilities for the AC600 PCCS. The unique feature of the PCCSAC code is the nine-equation model that classifies the fluid in the containment as steam, noncondensable air, and liquid water. The model includes the k-∊ turbulence model and a diffusion model for gas flow. The code considers nonuniform spatial factors. It can analyze most physical phenomena in the AC600 PCCS. The code was validated by comparing the present results with those of the COMMIX code for the analysis of a postulated small-scale model of AP600. The PCCS response of AC600 during design-basis accidents is analyzed.

Application of Some Turbulence Models to Simulate Buoyancy-Driven Flow

During a severe nuclear power plant (NPP) accident, large amounts of hydrogen and steam can be produced in nuclear reactor containment. In the case of hydrogen combustion, there is a possibility of producing short term pressure or detonation force. Therefore, these gas species’ production could threaten containment integrity. For instance, in the past, two gas explosion accidents occurred: In 1979 Three Mile Island and in 2011 Fukushima. After these accidents, modeling the gas behavior became an important topic in nuclear safety analyses. In order to predict hydrogen behavior and other gas species transport, mixing and combustion, reliable turbulence models need to be applied. In this work, standard k–e, k–ω, RNG k–e, Realizable k–e and SST k–ω turbulence models are addressed. The computations are performed with HYDRAGON code. HYDRAGON code is a three-dimensional thermal-hydraulic code, developed to solve low-speed gas flow of compressible Navier-Stokes equations in cartesian or cylindrical coordinates or a mixture of the two coordinates. The goal of this work is to test the performance of these models by comparing the results to the benchmark. The code aims to predict containment thermal-hydraulic conditions during NPP accident.Copyright

Assessment of Performance of Turbulence Models of CFX in Predicting Supercritical Water Heat Transfer in a Vertical Tube

Computational fluid dynamics (CFD) codes are widely used to investigate the thermal hydraulics of revolutionary reactor concepts in the last decade. Most CFD codes, including ANSYS CFX, solve Navier-Stokes equations by Reynolds averaging approach, which use turbulence models and wall functions. These turbulence models and wall function are validated against experimental data under subcritical pressures. However, the applicability of them to supercritical water heat transfer has to be examined. The objective of the present study is to evaluate the performance of the turbulence models of CFX in predicting supercritical heat transfer under non-buoyancy influence and strong buoyancy influence. The calculation results have been compared with two independent validation experimental data. The results indicate the need to improve a turbulence model to take into account the buoyancy effect and property variations on the turbulence for thermal-hydraulic calculations of the supercritical water when the buoyancy influence is strong.Copyright

The effect of turbulence modeling on hydrogen jet dispersion inside a compartment space using the HYDRAGON code

ABSTRACT The mitigation of hydrogen in the containment of nuclear reactor after the Loss of Coolant Accident is essential to preserve the structural reliability of the containment. This paper presents the results of the systematic work done by using the HYDRAGON code to investigate the effect of turbulence models on the concentration distribution of hydrogen and to determine the HYDRAGON code thermal-hydraulic simulation capability during a severe accident at the nuclear power plant. The HYDRAGON code is developed by the Department of Engineering Physics, Tsinghua University, which is an independent research program. The influence of various types of turbulence models, i.e. a standard k − ϵ model, a re-normalized group (RNG) k − ϵ model, and a realizable k − ϵ model were analyzed and the simulation results were compared with the experimental data. When simulation results were compared to experimental data, it was found that, in most compartments, the standard k − ϵ model generally yielded reasonable agreement with the experimental results as compared to RNG k − ϵ and realizable k − ϵ models; however, for probes P7 and P12, better trend was captured by RNG k − ϵ and realizable k − ϵ models, respectively.

Modeling the Irradiation Swelling of UO2 at the Fuel Pellet Rim

Irradiation swelling of UO2 at the fuel pellet rim was modeled based on the published theory and data of HBS (High Burnup Structure) formation. Fuel swelling was divided into two parts: fuel matrix swelling and porosity growth. Both solid fission products and fission gas contribute to the fuel matrix swelling prior to HBS transformation, resulting in relatively stable matrix swelling rate of around 1.0% per 10 GWd/tU, but the transformation accompanied by Xe depletion reduces the matrix swelling rate to approximately 0.3% per 10 GWd/tU, only attributed to solid fission products. Considering the direct impact of Xe depletion on the drop of matrix swelling rate, the exponential law of Xe depletion was applied to model the reduction of matrix swelling rate. Pore size and pore density evolution are the two main aspects of porosity growth. Pore size takes the form of lognormal distribution, whose parameters are obtained through fitting the experimental data. Pore density increases in the transformation process but goes down as a result of pore coarsening thereafter. Published data of three pellets were used to verify the correlations modeling pore growth, which were proven generally consistent with each other. The results of this work are ready to be incorporated into fuel performance modeling code as an option for detailed calculation of fuel swelling.Copyright

Probability Analysis of the Fuel Microsphere Distribution of Dispersion-Type Nuclear Fuel Plates

The mechanism of microsphere interconnection through cracks, which is observed in post-irradiation examination of dispersion-type fuel plates for research reactor use and is responsible for blistering, is analyzed. The accumulation of fission gas with burnup and the resulting increase of gas pressure cause the metal matrix to crack and the fuel plate to blister. Based on the experimental observation, a small cylinder in the fuel mini-plate is chosen as the simulation volume. Assuming the diameter of fuel microspheres is identical and in normal distribution separately and the position of fuel microspheres is in uniform random distribution in the matrix, the distribution of fuel volume fraction in the cylinder of the mini-plate is simulated using Monte Carlo method. The analysis shows that the distribution of the fuel volume fraction is in good agreement with the normal distribution. Mini-plates with different fuel volume fractions are further calculated for comparison. The calculation for a full-size plate shows that as for the fuel volume fraction 0.21 and the inventory of reactor core the existence of cylinders with a fuel volume fraction which is susceptible to blistering turns out a possible threat to the fuel reliability. Probability analysis proves to be an effective technique for the quantitative characterization of dispersions.Copyright

Development of a New Heat Transfer Model Near the Quench Front in the Reflooding Phase of a Tight Lattice

Heat transfer characteristics near the quench front in a reflooding process are quite complex. Large amount of vapor are generated, and the rod clad temperature drops rapidly to near saturation state. Until now, heat transfer mechanism in this region has not been well understood yet. Best estimate codes like RELAP5, COBRA-TF tend to treat the heat transfer mode near the quench front as transition boiling. However, when calculating the reflooding phase of tight lattice, these codes always under-predict the quench temperature, and also the slop of the temperature drop is predicted to be less steep than the experimental data. In this paper, a new heat transfer model near the quench front in the reflooding phase of a tight lattice is developed. Instead of transition boiling, transient liquid film evaporation is considered to be the main heat transfer mode in this region. It is supposed that heat released near the quench front is through liquid film evaporation. Through comparisons with experimental data, it can be concluded that the new model can better predict the quench temperature and the temperature drop slop.Copyright