Xiaoying Zhang’s research while affiliated with Sun Yat-sen University and other places

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Publications (20)


Transient 3D simulation for heating and melting process of PWR core after SBO
  • Article

May 2018

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21 Reads

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2 Citations

Annals of Nuclear Energy

Xiaoying Zhang

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Dekui Zhan

In the heat transfer between the core and coolant in a reactor pressure vessel (RPV), a bounding structure such as a shroud has an important influence on the heating and melting process of the core. To study the 3D transient heat transfer and melting process of the Hualong 1 PWR using an IVR strategy, a numerical analysis model has been established based on the 2D transient heat conduction of each rod, which is coupled considering the residual heat of the core, the convective and radiation heat transfer inside the core, and natural convection heat transfer of the outer water chamber. A transient 3D numerical simulation code, 3DTMCOR, has been developed to analyze the melting process of an IVR reactor core with high precision. The effects of heat generation from a zirconium–water reaction, radiation heat transfer, and the adoption of an IVR strategy on the heating and melting process of the reactor core have been investigated. It was found that the first melting time for the same 100 MW PWR from the 3DTMCOR code is about 2500 s later than that of MAAP. The core temperature will first decrease with a decrease in core power within 100 s, then increase as the water level drops down with a zirconium–water reaction of 100 s to 1000 s, then decrease again with a decrease in core power from 1000 s to 4300 s, and increase a second time when water in the RPV is evaporated after 4300 s. At 4860 s, a continuous melting area first appears in the core top. Radiation heat exchange is very important for transferring heat from the fuel rods to structures such as the control rods and bounding structures, which is also the main cooling mode in the core after water is evaporated. The first melting time with the precise radiation exchange model will occur 200 s later than in the results with the empirical model. The water chamber outside the RPV does not have a remarkable effect on the heating and melting process of the core.


Ablation and thermal stress analysis of RPV vessel under heating by core melt

March 2018

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21 Reads

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6 Citations

Nuclear Engineering and Design

Dekui Zhan

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Xiaoying Zhang

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[...]

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After a reactor core melts and collapses suddenly, the melting core accumulates on the base surface of the bottom head of a reactor pressure vessel (RPV), causing severe thermal ablation and thermal stress, endangering the safety of the RPV bottom head. In this study, a 1000-MW pressurized water reactor is considered an example to study the heat transfer ablation and thermal stress of an RPV lower head after a core collapses, by performing numerical simulations. A two-dimensional (2D) heat transfer model is used to analyze the coupling heat transfer between the wall surface of the RPV, two-layer melting corium pool, and outer water chamber. The transient 2D temperature and ablation of the lower head wall surface are calculated. The thermal stress of the RPV lower head and deformation are also investigated using ANSYS software. The results show that (1) The upper crust is the least thick, with a thickness of approximately 0.01 m, whereas the side crust is the thickest at approximately 0.12 m at the base of the lower head. (2) The minimum thickness of the bottom wall decreases linearly with time, starting from the collapse time of the core to 2500 s, when it becomes 0.04 m. It does not change thereafter, but the melting zone is further expanded. (3) The lower head wall starts to melt from 200 s after the core collapses. The melting mass first increases sharply, and subsequently, increases slightly with time. The total melted material is 3000 kg at 5000 s. (4) The heat flux at the inner and outer surfaces of the lower head immersed in the uranium melt layer increases with the azimuth angle, reaching a maximum value of 750 kw/m² and 250 kw/m², respectively, at the interface with the metal melt. The heat flux at the RPV inner wall covered by the metal melt is approximately constant at 400 kw/m², whereas that at the outer surface decreases with the azimuth angle. (5) The stress at the lower head is concentrated at the inner surface, with a maximum value of 625.65 MPa. The radial deformation increases with time only until 2200 s, with a maximum deformation of 28.39 mm occurring in the lower part of the RPV bottom.


Simulations for cooling effect of PCCS in hot leg SB-LOCA of 1000 MW PWR

August 2017

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36 Reads

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1 Citation

Nuclear Engineering and Design

The passive containment cooling system (PCCS) for reactor containment is a security system that can be used to cool the atmosphere and reduce pressure inside the containment in case of temperature and pressure increase caused by vapor injection, which requires no external power since it works only with natural forces. This article aims to establish a CFD simulation model for the Passive Containment Cooling System of 1000 MW PWR using Code_Saturne and FLUENT software. The comparison of 4 different models based respectively on mixture model, COPAIN test, Uchida correlation and Chilton-Colburn analogy are used to simulate the condensing effect, the improvement of source code are based on a 3D simulation of PCCS system. To simulate the thermal-hydraulic condition in the containment after LOCA accident caused by a double-ended main pipe rupture, a high temperature vapor with the given mass flow rate are supposed to be the source of energy and mass. Meanwhile the surface of three condensing island use the wall condensation model. The simulation results obtained from the 4 models show similar transient process, the difference between the steady-state analysis of three models is less than 3%. Meanwhile, the large mass flow rate of water loss inside the containment induce a high flow rate of steam which could be uniformly mixed with air in a short time, the impact of steam injection to different directions on the simulation results is not obvious. For the self-condensing efficiency of 3 groups of PCCS system, the non-centrosymmetric injection position resulting that the condensing efficiency is slightly higher for the two heat exchanger groups nearby. During the first 2400 s of simulation time, more than 75.69% of the steam is condensed, the occurrence of condensation at the wall mainly driven by natural convection, the effect of thermodynamic siphon could improve the flow rate of gas mixture inside the tubes when the velocity of mixture is not large enough. The steam could smoothly enter the tube and reach the internal cooling surface then to be condensed. Besides, PCCS ensure that the containment internal pressure is maintained below 2 bar and the temperature is maintained below 380 K during 3600 s.


CFD Simulation of Passive Containment Cooling System in Hot Leg SB-LOCA for 1000MW PWR

July 2017

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21 Reads

The passive cooling system (PCCS) for reactor containment is a security system that can be used to cool the atmosphere and reduce pressure inside of containment in case of temperature and pressure increase caused by vapor injection, which requires no external power because it works only with natural forces. However, as the driving forces from natural physical phenomena are of low amplitude, uncertainties and instabilities in the physical process can cause failure of the system. This article aims to establish a CFD simulation model for the Passive Containment Cooling System of 1000MW PWR using Code_Saturne and FLUENT software. The comparison of 4 different models based respectively on mixture model, COPAIN test, Uchida correlation and Chilton-Colburn analogy which simulate the condensing effect and the improvement of source code are based on a 3D simulation of PCCS system. To simulate the thermal-hydraulic condition in the containment after LOCA accident caused by a double-ended main pipe rupture, a high temperature vapor with the given mass flow rate are supposed to be the source of energy and mass into containment. Meanwhile the surface of three condensing island applies the wall condensation model. The simulation results show similar transient process obtained with the 4 models, while the difference between the transient simulation and the steady-state analysis of three models is less than 3%. The large mass flow rate of water loss status inside the containment cause a high flow rate of vapor which could be uniformly mixed with air in a short time. For the self-condensing efficiency of 3 groups of PCCS system, the non-centrosymmetric injection position resulting that the condensing efficiency is slightly higher for the two heat exchanger groups nearby. During the first 2400s of simulation time, more than 75.69% of the vapor is condensed, indicating that for the occurrence of condensation at the wall mainly driven by natural convection, the effect of thermodynamic siphon could improve the flow of gas mixture inside the tubes when the velocity of mixture is not large enough, so that the vapor could smoothly enter the tube and reach the internal cooling surface then to be condensed. Besides, PCCS ensure the containment internal pressure maintained below 2 bar and the temperature maintained below 380K during 3600s.


Numerical Simulation of Core Temperature and Melting Process of IVR Core After a Severe Water Loss Accident

July 2017

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24 Reads

After the occurrence of severe water loss accident in a PWR, the water level in the reactor core would decrease gradually, leading to heat up and melted down of the core, threatening safety of the nuclear power plant and the surrounding environment. In this paper, the 1/4 core of AP1000 PWR was adopted for study, a numerical method has been established to calculate the transient change of temperature and melting process of the core and envelope structure (boarding, basket and RPV) after the severe water loss accident. A two-dimensional conduction model with cylindrical coordinate has been used to simulate heat transfer along the radius and height direction of fuel rods and control rods in fuel assemblies. Heat transfer condition on rod surface considers nucleate boiling for rod surface below the water level, while radiative heat transfer among neighboring rods and natural convection with water vapor was considered for rod surface above the water level. Heat transfer along thickness of envelope structures were modeled with the one-dimensional conduction model. The results show that the maximum temperature of the whole reactor core does not exceed 3000K and AP1000 will not meet the melting of fuel rods with the help of RPV external water chamber cooling. The temperature values of the fuel rods and the control rod show the characteristic distribution of the two regions. At 4904s, the maximum temperature of the rod rises to 2900K, and then stabilize. The temperature of the shell is up to 2000K, the maximum temperature of the basket is to 1260K, the variation of RPV wall temperature is not obvious.


Transient Simulation on Reactor Core Melt and Lower Support Plate Ablation in In-Vessel Retention

July 2017

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7 Reads

To precisely understand the accident process of reactor core melt in In-vessel retention (IVR) condition, 3-dimensional transient thermal conduction analysis with moving boundary is performed on quarter reactor core model. The decline of decay power and water level in reactor pressure vessel (RPV), and the radial distribution of assemblies of different material is considered. Convective heat transfer on rod surface and coolant interface is computed with empirical correlation of natural convection of saturated steam vapor / water. Radiation heat transfer with 16 neighboring rod is considered. Also, a dynamic ablation model is developed to simulate the ablation of lower support plate (LSP) caused by continuously accumulation of molten corium. The impingement heat transfer of the falling corium and the molten pool formed in LSP ablation cavity is taken into account. The simulation gives the ablation process on the surface of LSP as well as temperature history and molten proportion of the reactor core, which shows agreement with reference. Simulation shows: the melt process of reactor core accelerated in the accident process of 2600s, when coolant in RPV dry up. 65% of the core mass has molten at 8000 second. LSP is completely penetrated in 6000s, the ablation of LSP is mainly focused on an annular region of radius 700mm.


A two-dimensional experimental investigation on debris bed formation behavior

April 2017

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24 Reads

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20 Citations

Progress in Nuclear Energy

Studies on debris bed formation behavior are of crucial importance for the improved evaluation of Core Disruptive Accidents (CDA) that could occur for Sodium-cooled Fast Reactors (SFR). In this work, to clarify the mechanisms underlying this behavior, a series of experiments was performed by discharging solid particles into two-dimensional rectangular water pools. To obtain a comprehensive understanding, various experimental parameters, including particle size (0.256∼8 mm), particle density (glass, alumina, zirconia, steel and lead), particle shape (spherical and irregularly-shaped), water depth (0–60 cm), particle release pipe diameter (10–30 mm), particle release height (110–130 cm) as well as the gap thickness of water tank (30–60 mm), were varied. It is found that due to the different interaction mechanisms between solid particles and water pool, four kinds of regimes, termed respectively as the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertial dominant regime, are identified. The performed analyses in this work also suggest that under present experimental conditions, the particle size, particle density, particle shape, particle release pipe diameter and water depth are observable to have remarkable impact on the above regimes, while the role of particle release height and gap thickness of water tank seems to be less prominent. Knowledge and data from this work might be utilized for the improved design of core catcher as well as analyses and verifications of SFR severe accident analysis codes in China in the future.


CFD analysis of flow field in a 5 × 5 rod bundle with multi-grid

October 2016

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76 Reads

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42 Citations

Annals of Nuclear Energy

In order to understand the characteristics of flow field in a fuel bundle with multi-grid, in this paper detailed CFD analyses were performed against a 5 × 5 rod bundle with two spacer grids and one mid span mixing grid. To represent the actual grid as accurately as possible, in our calculation the grids are described as a structure of stripe, dimple, spring, and mixing vane. Based on the calculated results, characteristics of flow field at different channels along the length of fuel assembly, flow information at important specific sites as well as the influence of mixing vane geometry (namely its deflection angle and vane length) are investigated and compared. It is shown that the upstream grid has no remarkable impact on the flow field formed at its downstream grids if the cross flow is sufficiently developed. Both the spacer grid and mid span mixing grid are verifiable to play an important role in changing the flow behavior and enhancing the heat transfer among the bundle channels. In addition to the mixing vane, the structure of dimple and spring is also observable to have some impact on the local flow field at the grids. The performed analyses in this work suggest that increasing the deflection angle and length of mixing vanes will lead to much strengthened mixing performance and enhance the local heat transfer, along with an appropriately increased pressure drop simultaneously.


Fig. 1 Flow data of plume from the full-sized rocket motor.  
Fig. 4 Calculated radiance with Monte-Carlo method by Burt and Boyd. 22  
Fig. 5 Global radiance computed with code of the work in this paper.  
Fig. 6 Plume radiance of reduce-scale and full-sized rocket motorsat normal direction in 2.7 lm.
Fig. 7 Plume radiance at normal direction in 4.2 lm of reduced-scale and full-size rocket motors.  

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Numerical study on similarity of plume infrared radiation between reduced-scale solid rocket motors
  • Article
  • Full-text available

June 2016

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454 Reads

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12 Citations

Chinese Journal of Aeronautics

This study seeks to determine the similarities in plume radiation between reduced and full-scale solid rocket models in ground test conditions through investigation of flow and radiation for a series of scale ratios ranging from 0.1 to 1. The radiative transfer equation (RTE) considering gas and particle radiation in a non-uniform plume has been adopted and solved by the finite volume method (FVM) to compute the three dimensional, spectral and directional radiation of a plume in the infrared waveband 2-6 μm. Conditions at wavelengths 2.7 μm and 4.3 μm are discussed in detail, and ratios of plume radiation for reduced-scale through full-scale models are examined. This work shows that, with increasing scale ratio of a computed rocket motor, area of the high-temperature core increases as a 2 power function of the scale ratio, and the radiation intensity of the plume increases with 2-2.5 power of the scale ratio. The infrared radiation of plume gases shows a strong spectral dependency, while that of Al2O3 particles shows spectral continuity of grey media. Spectral radiation intensity of a computed solid rocket plume’s high temperature core increases significantly in peak radiation spectra of plume gases CO and CO2. Al2O3 particles are the major radiation component in a rocket plume. There is good similarity between contours of plume spectral radiance from different scale models of computed rockets, and there are two peak spectra of radiation intensity at wavebands 2.7-3.0 μm and 4.2-4.6 μm. Directed radiation intensity of the entire plume volume will rise with increasing elevation angle.

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A similarity study on the infrared radiation of solid rocket plume in different reduced-scale sizes

June 2015

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18 Reads

IEEE Aerospace Conference Proceedings

The similarity of the plume radiation of solid rocket engines in different reduced-scale sizes is important for studying the radiation of regular-sized engines with small-scale tests. To determine the similarity of plume radiation between reduced-scale rockets and regular-sized rockets in ground-test conditions, the flow and radiation of rocket engines in the geometric reduced-scale ratios of 0.1 to 1 are investigated in this study. First, a Calculate Fluid Dynamic (CFD) solver is used to obtain the flow field of the internal flow and plume. Thereafter, the Finite Volume Method (FVM) is applied to compute the 3D spectral-directed plume radiation in the 2- 6μm infrared waveband for the nonuniform absorptive/ emissive/scattering plume. The spectrum characteristics of gaseous compositions and Al2O3 particles are solved by using the weighted-sum-of-gray-gas (WSGG) model and Mie theory, respectively. Finally, the integral radiation on the surface of a high-temperature plume core (higher than 500K) that produces the volume radiation in different directions was investigated. The direction of the volume radiation is determined by the angle between the radiation direction and rocket axis. Our research shows that with the decreasing size of the rocket engine, the high-temperature core's area decreases with the square order of the rocket size. The infrared spectral radiation of the plume also decreases with the square order. The infrared radiation of the gaseous components show a strong spectral difference, and the infrared radiation of the Al2O3 particles show the spectral property of a gray medium with the same temperature. The integrated infrared characteristics of the solid rocket plume mainly show the spectral continuity of Al2O3 particles, which decreases in the peak radiation spectrum of gaseous components. The emission and scattering of Al2O3 particles makes the plume radiation grow up remarkably, this phenomenon increases the plume radiation in the 4.2-4.45μm band to two times of the nonparticle radiation and increases the plume radiation in the 2.7-2.95μm band by 45%. The radiation intensity on the surface of the high temperature plume core increases with increasing sight angle.


Citations (9)


... The first investigated cycle was the supercritical Rankine cycle, chosen because of the advanced studies on the technology of its components. An energy balance was performed on the preliminary scheme presented in Figure 2 and based on the cycle reported in [16]. The thermal power coming from the intermediate circuit is exchanged through the secondary steam generator, which allows the secondary feedwater entering at 320 • C to become supercritical steam at 540 • C, at a pressure of 250 bar. ...

Reference:

Analysis of Power Conversion System Options for ARC-like Tokamak Fusion Reactor Balance of Plant
Control and thermal analysis for SCWR startup
  • Citing Article
  • December 2019

Annals of Nuclear Energy

... In most of integral reactor designs, reactor core, coolant pumps, steam generators and pressurizer are all packed inside the reactor pressure vessel (Buongiorno et al., 2012;Fetterman et al., 2011;Wu, et al., 2016;Wang et al., 2020aWang et al., , 2020b to form a compact system, as shown in Fig. 1. Such configuration makes it possible for pre-fabrication and testing at factories, which simplifies on-site installation process, reduces the number of auxiliary systems, and eliminates many pipe connections, thus, reducing the potential risks sources of loss of coolant accidents. ...

Design and analysis of improved two-phase natural circulation systems with thermoelectric generator
  • Citing Article
  • May 2020

Annals of Nuclear Energy

... The program embedded heat transfer and pressure-drop relationships, and the results demonstrated good agreement with RELAP5 calculation. Chen et al. (2019) developed the THOSG code for thermal-hydraulic analysis, simulating the thermal-hydraulic behavior of OTSG by coupling flow and heat transfer on the primary and tube sides. Yao et al. (2021) developed the code TACS, utilizing a homogeneous flow model for simulating two-phase flow in helically coiled tubes. ...

Development of thermal-hydraulic analysis code of a helically coiled once-through steam generator based on two-fluid model
  • Citing Article
  • October 2019

Annals of Nuclear Energy

... Modern works of Daude et al. [15,16] and Delchini et al. [17] on two-phase flows also use one-dimensional systems of balance equations and are devoted to developing a method for simulating non-stationary flows in variable geometry pipes. In the work of Chen et al. [18], the basic system of equations was transformed into a new form to obtain a coupling matrix equation. As a result, instead of a multi-step solution process, a one-step solution was obtained. ...

Implementation and validation of a one-step coupled solution method for the two-fluid model
  • Citing Article
  • July 2019

Nuclear Engineering and Design

... In most cases, pressurized water is used as the coolant in these types of reactors, and the classical convective heat transfer correlations can be applied for calculations [2]. However, the literature review results [2][3][4][5][6][7][8][9][10][11][12][13][14][15][16] show that there is increased interest in nanofluid application for heat transfer enhancement in PWRs. Section 3 reports a review of convective heat transfer correlations valid for BWRs. ...

Transient 3D simulation for heating and melting process of PWR core after SBO
  • Citing Article
  • May 2018

Annals of Nuclear Energy

... In this work, the geometry and boundary conditions are viewed as an axisymmetric model as shown in Fig. 19a. The molten core can be simplified to a non-uniform heat flux q h ð Þ which is shown in Fig. 19b (Zhan et al. 2018). The initial temperature of entire structure of RPV is set to be T 0 ¼ 127 C and the temperature of the outer wall is assumed to be T b ¼ 127 C. For the mechanical boundary conditions, the top surface of RPV is fixed in the z direction. ...

Ablation and thermal stress analysis of RPV vessel under heating by core melt
  • Citing Article
  • March 2018

Nuclear Engineering and Design

... The characteristics of the criticality of the fuel debris are expected to vary during its dispersion and sedimentation in DBF process (Muramoto et al., 2019(Muramoto et al., , 2021Shiba et al., 2023). It was also found that the formed debris beds may perform different geometrical characteristics at different hypothetical accident conditions (e.g., debris properties, coolant boiling) due to the varying multi-phase interactions (Lin et al., 2017;Cheng et al., 2018c; σ surface tension (N⋅m − 1 ) ϕ particle sphericity (-) ψ quantity characterizing the bubbling effect on restraining the particle-flow induced pool convection (-) Ω convergence degree of mixed-sized particle mixture (-) Subscripts a area mean term B boiling b particle bed c critical value convection pool convection cr cross section dim dimple eff effective ev equivalent f fluid g gas in inner container inertia particle inertia ip initial value of particle L left l liquid level fully leveled off m particle mound n nozzle p particle pr particle releasing R right r repose re relative quantity tank water tank top topmost particle tp two-phase fluid v stands for a volume mean term w water (Cheng et al., 2013c;Suzuki et al., 2014;Lin et al., 2017). ...

A two-dimensional experimental investigation on debris bed formation behavior
  • Citing Article
  • April 2017

Progress in Nuclear Energy

... In nuclear engineering, thermal-hydraulic analysis of the reactor core focuses particularly on the fully developed flow characteristics within the rod bundle subchannels. (Bae and Park, 2011;Cheng et al., 2017;Ma et al., 2022;Podila and Rao, 2016;Wang et al., 2020;Zhou et al., 2015;Zimmermann et al., 2015). A nuclear reactor core is composed of tens of thousands of fuel rods. ...

CFD analysis of flow field in a 5 × 5 rod bundle with multi-grid
  • Citing Article
  • October 2016

Annals of Nuclear Energy

... Kim et al. [7] experimental results show that the different sizes of the rocket motor of the same propellant, infrared radiance and the spectral characteristics of the combustion products are very similar, indicating a relationship between the infrared emission properties of tiny standard rocket motors and actual rocket motors. Zhang et al. [8] determined that the nozzle radius of a Trident D5 solid rocket motor should be scaled to 0.1 to 1 times the full-size nozzle. They discovered that the ratio of radiance to radius follows an exponential function within a range of 1.5 to 2.5. ...

Numerical study on similarity of plume infrared radiation between reduced-scale solid rocket motors

Chinese Journal of Aeronautics