External Reactor Vessel Cooling Research Articles

Study on the lower head failure in severe accidents Part II: Thermal-mechanical coupling behavior analysis on CAP1400 reactor lower head

Under severe accident conditions with core meltdown, large amounts of molten material relocate to the lower head at high temperatures. Significant thermal and mechanical loads may lead to failure of the reactor pressure vessel (RPV) lower head, spreading radioactive fission products. Lower head failure time and location are crucial for accident mitigation and consequence assessment. This study developed a thermodynamic failure analysis model for the lower head by investigating material failure mechanisms, coupling this into an Integrated Severe Accident Analysis code (ISAA), and conducting experimental validation and applied research. Part I mainly introduces the thermodynamic analysis module development, validation against OLHF experiments, and preliminary failure criteria analysis. Results indicate the model accurately predicts lower head failure behavior, and Larson-Miller life fraction and Hedl-Dorn parameter criteria accurately assess lower head integrity. Part II presents applied research after validation and preliminary analysis. Using the improved ISAA, melt pool behavior and lower head thermal–mechanical response were analyzed for a CAP1400 small break loss-of-coolant accident (SBLOCA), assuming multiple safety system failures. Results indicate the bottom of the lower head melt pool didn’t form a hard shell due to high heat release rate, only an oxide shell layer around the second layer. Decay heat from melt pool components cannot be dissipated, leading to lower head failure. Further External Reactor Vessel Cooling (ERVC) heat transfer model analysis and research are needed in ISAA. The two developed mechanical analysis models predict material failure at high temperatures, and the timing and location of lower head failure are consistent, at approximately 16,000 s at 25.38°±5.47° on the lower head. The predicted failure mode is similar to Part I validation, indicating both Larson-Miller life fraction and Hedl-Dorn parameter criteria provide consistent lower head integrity judgments.

Numerical simulation study of boiling Critical Heat Flux characteristics of graphene nanofluids

IVR-ERVC (In-Vessel Retention – External Reactor Vessel Cooling) is a critical accident management method for ensuring the integrity of the reactor pressure vessel (RPV) lower head. One of the most crucial aspects within this method is to enhance the CHF (Critical Heat Flux) on the outer surface of the reactor pressure vessel (RPV) lower head. This paper explores the application of graphene nanofluids in IVR-ERVC. This paper uses computational fluid dynamics to numerically simulate the CHF characteristics of graphene nanofluids under different undercooling, mass flow rate, concentration, tilt angle, and gap size conditions and analyzes the impact of different factors on CHF and the coupling effects between different factors. The undercooling range is 5 K–100 K, the mass flow rate range is 150 kg/(m2∙s) to 3000 kg/(m2∙s), the concentration range is 0–1%, the inclination angle range is 0°–90°, and the gap sizes are 10 mm and 20 mm. The results show that the CHF can be effectively improved by adding nano-graphene into the base solution, and the CHF increases with the increase of subcooling degree, mass flow rate, concentration, dip Angle and gap size. Within the simulation range, the strengthening effect on CHF weakens when raising the undercooling and mass flow rate. And the larger the concentration and inclination angle within the simulation range, the better the enhancement effect on CHF. The maximum increase was 290%, while the average increase was 75.6%.This article studies the coupling effects between different parameters through numerical simulation. It can be concluded that increasing the mass flow rate weakens the enhancement of subcooling and concentration on CHF, increasing the subcooling weakens the enhancement effect of mass flow rate and concentration on CHF, and increasing the concentration enhances the enhancement effect of subcooling and mass flow rate on CHF. Enhance the mass flow rate will weaken tilt angle effect, while increasing the concentration will enhance the strengthening effect of tilt angle on CHF.

A critical heat flux model for IVR-ERVC application based on the experimental observation

It is essential to conduct theoretical and experimental research on the heat transfer limit for External Reactor Vessel Cooling (ERVC), which aims to implement In-Vessel Retention (IVR) severe accident mitigation strategy successfully. In this paper, a theoretical model for IVR-ERVC condition is proposed to predict the critical heat flux (CHF) based on the experimental observation. According to the experimental observation on the one-dimensional full height IVR-ERVC experimental facility, it is found that there is a thin liquid film between the heating surface and the vapor slug. And this film is evaporated completely when CHF occurs. In addition, a large number of bubble condensation collapses were observed under subcooling condition. According to these experimental observations, a CHF model based on the mechanism of liquid film dryout is developed. The CHF trigger mechanism of this model is that the supply of fresh liquid into the liquid film is not sufficient to prevent depletion of the liquid film due to the boiling. The influence of heating surface geometry and the heat flux shape are considered in this model. The predicted values of the model are compared with the CHF experimental result from different test facilities with different conditions, and the comparison result shows that the relative errors are basically within 20 %. The analysis results of the model indicate the increasing of coolant subcooled degree is conductive to the improvement of CHF. When the coolant subcooled degree increases by 15 ℃, more than 30 % maximum increase in CHF is obtained. In addition, this model can reflect the “exit phenomena” to a certain extent.

Uncertainty analysis of molten corium In-Vessel Retention under External Reactor Vessel Cooling

External Reactor Vessel Cooling (ERVC) is a key measure to realize molten corium In-Vessel Retention (IVR). The uncertainty of severe accident in core leads to the uncertainty of molten pool parameters in lower head, which may threat the reliability of successful implementation of IVR. In this study, an uncertainty analysis is employed to ensure the reliability of ERVC technology to IVR with the final two-layer molten pool configuration of AP1000.600 simulations are performed to determine the effect of five key parameters. As long as parameters are within high probability density range, lower head can remain intact, and safety margin is greater than 15 %. For influence level and correlation of parameters, the order on metal layer is: Decay heat > Fe mass > UO2 mass > Zr oxidation fraction > Emissivity of metal layer. The results can provide guidance for subsequent research direction and optimal design of mitigation measures.

Numerical study of thermal-hydraulic and structural parameters effects on CHF in subcooled forced convection flow during external reactor vessel cooling

Advanced nuclear reactors employ External Reactor Vessel Cooling (ERVC) as a crucial measure to safeguard the structural integrity of the Reactor Pressure Vessel's (RPV) lower head and to confine molten material within it during severe accidents. The ERVC technique effectively mitigates severe accidents by efficiently dissipating decay heat from the molten pool located within the lower head, utilizing coolant flow boiling. The key determinant of the upper limit of ERVC cooling capability is primarily contingent on the Critical Heat Flux (CHF). In this study, flow boiling in downward heating curved flow channel under forced convection is numerically investigated using the Eulerian-Eulerian two-fluid model and realizable k-ε turbulence model. Compared with the experiment, the maximum deviation of numerically predicted CHF is less than 5%, indicating that the numerical method is reliable. A thorough analysis is conducted on the impact of variable structure and thermal-hydraulic operating conditions on CHF. The results indicate that as the coolant mass flow rate, system pressure, and inlet subcooling increase, the CHF exhibits a corresponding increase. The CHF increases by 56.6% when the mass flow rate increases from 400 to 1000 kg/(m2·s). The CHF at the pressure of 0.3 MPa is 16% higher than that of 0.1 MPa. The CHF increases from 1.14 to 1.84 MW/m2 as the inlet subcooling escalates from 10 to 20 K, with a 60% growth rate, while the CHF increases to 2.03 MW/m2 when the inlet subcooling increases from 20 to 50 K, with a 10.33% increase. The maximum difference in CHF for the three flow channel widths (120, 150, and 170 mm) is only 1.7%, indicating that the influence of flow channel widths is insignificant. This study can provide theoretical guidance for the design and optimization of ERVC in advanced nuclear reactors.

Critical heat flux and heat transfer behaviors on aged RPV outer wall surface in flow boiling under IVR conditions

Critical Heat Flux (CHF) plays an important role in the nuclear power plant safety strategy both during normal operation and under accident conditions. During normal operation, the outside wall is exposed to high-temperature environment. The surface is oxidized by air, which is so-called “aged surface”. To specify the characteristics of aged surface and the effect on CHF, the specimen aging experiments and flow boiling experiments under External Reactor Vessel Cooling (ERVC) were conducted. The results of specimen aging experiments show that surface characteristics get steady after 500 h aging and the oxide layer is the Fe-Fe2O3-Fe3O4 mixture. After aging, the surface gets rougher and has better hydrophily. In the flow boiling experiments, characteristics of heat transfer and CHF are studied. Heat transfer is enhanced by extended surface area and increased nucleation sites at lower angle. The influence of capillary suction is not obvious and the enhancement of CHF is limited. As the inclination angle increases, the departure of bubbles is more frequent and capillary suction is dominant to assist liquid supply into the liquid sublayer and delay the boiling process. At higher angle, CHF is more obviously enhanced but heat transfer gets worse. Overall, CHF of RPV outer wall is increased after surface aging, and it is beneficial to improving the safety margin of IVR mitigation measures.

Enhancement of the critical heat flux for downward-facing saturated pool boiling on the reticular hollow shell structure surfaces

In-vessel retention (IVR) technology was currently an effective strategy for retaining molten materials in the reactor pressure vessel (RPV) when the nuclear power plants occurred severe accident, further ensuring the integrity of RPV. The effectiveness of IVR depended on the critical heat flux (CHF) of external reactor vessel cooling (ERVC). A novel type of reticular hollow shell structure (RHS) was designed to meet this need. Five RHS surfaces were designed and the pool boiling heat transfer performance were investigated in saturated deionized water, the experimental inclination angles were 5°, 30°, 45°, 60° and 90°. The boiling heat transfer coefficient (BHTC) and CHF were measured by steady-state heating and transient quenching methods. The experiment results showed that the CHF increased with the increase of inclination angles, and the CHF of all RHS surfaces were significantly enhanced compared with the plain surface. The maximum CHF value and the maximum CHF enhancement exhibited on surface 4, with a maximum CHF of 2992.2 kW/m2.When the inclination angle was 30°, the maximum CHF could reach 2338.6 kW/m2, which was 160.3% enhancement over the plain surface. Surface 1 and surface 3 had better BHTCs, and the maximum values were 178.3 kW/(m2·K) and 126.8 kW/(m2·K), respectively, which were much better than other structured surfaces. The RHS could form a gas–liquid conversion system, which was composed of cavities and alternating channels to increase the surface nucleation sitesand accelerate the circulation of cooling water, thus obtained the high CHF and high BHTC characteristics. RHS provided a large enough CHF margin to ensure that the RPV was continuously cooled and its integrity was maintained. Finally, the behavior of bubbles generation, coalescence, sliding and detachment in reticular hollow shell structure were systematically analyzed.

Transient heat transfer and crust evolution during debris bed melting process in the hypothetical severe accident of HPR1000

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m2. The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20–50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

Study on the safety characteristics of molten corium in-vessel retention of annular fuel core

In the process of reactor severe accidents, maintaining the integrity of reactor pressure vessel to achieve IVR (In-Vessel Retention) of molten corium plays an important role in accidents mitigation and suppression. The ERVC (External Reactor Vessel Cooling) is an important technology to realize IVR. As a new economical annular fuel, the increase of zirconium content and power density in the core may threaten the integrity of lower head under severe accidents. In this paper, the IVRASA (In-Vessel Retention Analysis code in Severe Accident) is used to analyze the heat transfer characteristics of lower head molten pool corresponding to annular fuel core. The components of two-layer and three-layer configurations of molten pool are determined based on the geometric structures of annular fuel and verified heavy metal layer model. In terms of the integrity of lower head, the power density of the annular fuel core can be increased to about 110%.

Numerical Investigation on the Thermal-Hydraulic Characteristics near a Corroded Lower Head of the Reactor Vessel During IVR-ERVC

Third-generation advanced pressurized water reactors adopt the external reactor vessel cooling (ERVC) strategy to ensure the pressure vessel is not at risk of melt-through in severe accidents, thereby completely controlling radioactive materials in the pile. However, due to its long-term service, vessel aging, steel corrosion, and oxidation may lead to deformation at different locations on its outer surface, forming various shapes of sawtooth structures, thus affecting the heat transfer behavior of the high-temperature walls during ERVC. In this paper, the Fluent code, coupled with boiling heat and mass transfer equations based on user-defined functions (UDFs) was used to simulate the thermal-hydraulic processes on the lower head with three typical sawtooth structures. The distribution of the stagnation zone for vapor buildup was the main focus. By varying the heat flux installed on the lower head and the inlet velocity of the flow channel, the onset time of critical boiling and the development of the location of critical heat flux over time were further investigated. It was found that a long period of unstable bubble generation and detachment before the onset of critical boiling may occur on the lower head. These findings can provide technical support for the safety design of advanced nuclear reactors.

Study on structural integrity of reactor pressure vessel with various ablation under core meltdown accident

In-vessel retention is an important measure to alleviate the core meltdown accident in the third generation nuclear power plant. External Reactor Vessel Cooling(ERVC) is the key to the successful implementation of IVR. Due to the ablation caused by the molten pool in the lower head, the reactor pressure vessel(RPV) has various wall ablation and thinning, which poses a great threat to the integrity of the pressure boundary of the RPV. Based on the boiling water RPV as the research object, the model is established for the RPV with various ablation and local thinning by finite element method(FEM). The stress/strain response of the structure is calculated and analyzed under the core meltdown accident. The plastic and creep damage are assessed for the RPV in terms of its evolution and distribution with continuum damage mechanics(CDM). Due to the existence of internal pressure within the RPV, the effect of pressure on creep and plastic rupture is also studied within the required 72 h. According to the criterion of ductile exhaustion, the results show that the increase of internal pressure further reduces the load-bearing capacity of RPV.

Effect of metal layer height on heat transfer inside molten pool

Abstract In a serious accident, after the core of a nuclear reactor melts and collapses, In-Vessel Retention in the External Reactor Vessel Cooling (IVR-ERVC) is an effective technology to maintain the integrity of lower head by reducing heat load on it. The various factors affecting the melting of the lower head have been widely studied. The mass of the molten metal layer may affect the consequences of the accident, since it is where the focusing effect occurs. However, the related research is still absent. In this paper, we systematically calculated the heat transfer behavior and melting process under different metal layer heights conditions. The temperature distribution, the velocity distribution, the heat flux of the outer wall of the Reactor Pressure Vessel (RPV), and the change of the thickness of the RPV were obtained through Large Eddy Simulation (LES). Interestingly, the heat flux increases with the metal layer height at first and achieve the maximum in the middle height. These results increase the understanding towards the serious accidents.

External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krško

The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krško applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant’s site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant’s post-Fukushima improvements and adequacy with plant’s procedures for severe accident mitigation. Conclusion is that NPP Krško could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance’s (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the lost core geometry.

Evaluation of the success of in-vessel corium retention using external reactor vessel cooling mechanism for a small PWR

Among the several study fields of severe accidents, In-Vessel Retention (IVR) via External Reactor Vessel Cooling (ERVC) referred to as IVR-ERVC is a critical subject being explored for various reactor designs. This study aims to investigate and assess the effectiveness of IVR of molten corium in a PWR utilizing the ERVC mechanism. A RELAP5/SCDAP model developed for the assessment of IVR-ERVC was applied to a small PWR after its successful validation. The boundary conditions applied to the model are evaluated through simulation of severe accidents of the small reactor (998 MWth). Thermal hydraulic studies of sensitive parameters have been carried out to determine the safety margin and effectiveness of the IVR-ERVC method under a variety of process conditions. The results indicate that a high mass flux is required to avoid boiling crises at low degrees of subcooling (around saturation conditions). A mass flux of 186.44 kg/m2-s or higher is required to provide a sufficient safety margin for inlet water with degree of subcooling of 8 K. The minimum Critical Heat Flux (CHF) is found to be 2.5 MW/m2 at a mass flux of 186.0 kg/m2-s for this case. Critical Heat Flux Ratio (CHFR) is found in the range of 6.0–17.0 with the cooling water flow rate of 20 kg/s (mass flux of 34.48 kg/m2-s) and 65 K subcooling, which provides a sufficient safety margin to avoid boiling crisis. It is concluded that the ERVC is an effective mechanism for the prevention of RPV rupture under active and passive conditions of the reference reactor.

Transient behavior and maximum heat flux ratios of two-layer corium pool

In this study, the transient behavior of two-layer corium pool was studied based on COPRA facility. The transient change of two-layer corium pool caused by power change and the effect of stratification and stratified plate thickness on heat transfer characteristics were investigated. Then maximum heat flux ratios (MHFRs) of two-layer corium pools was proposed and investigated using CFD code FLUENT. The heat flux ratio is the ratio of heat flux to qCHF of vessel wall. The maximum of heat flux ratios at all angles is the MHFR of the corium pool. The MHFRs of two-layer corium pool under radiation top surface condition and isothermal top surface condition were obtained. By comparing the values and positions of the MHFRs, the degree of approaching boiling crisis of local reactor pressure vessel wall can be quantitatively analyzed. The results show that MHFR appears at metallic layer wall under radiation top surface condition and at upper part of the oxide layer wall under isothermal top surface condition. The results of this research can provide references for IVR (In-Vessel Retention) safety analysis and ERVC (External-Reactor Vessel Cooling) methods such as Nano-coating measure.

Machine learning prediction of critical heat flux on downward facing surfaces

Application of external reactor vessel cooling (ERVC) in-vessel retention (IVR) was an important measure to ensure the integrity of the lower head of a reactor pressure vessel (RPV). As a typical boiling heat transfer process, the critical heat flux (CHF) plays a crucial role in the safety margin of a RPV's IVR-ERVC strategy. Although there have been a lot of correlations and experiments about the CHF of pool boiling on downward facing surfaces, their application range was relatively limited. To further expand the usability, machine learning was introduced in this research after collecting most accessible CHF data of pool boiling on downward facing surfaces. Considering the small amount of these experimental data, some pseudo data obtained by fitting the existed experimental data were added. Three machine learning methods, the ε-support vector machine (ε-SVM), back propagation neural network (BPNN) and random forest were used to predict CHF. Among the three methods, ε-SVM provided the best accuracy in the prediction. The effects of orientation, surface dimensions, heating surface material and pressure on the downward facing surface boiling crisis were predicted by ε-SVM, the results showed that in general, the CHF increased with the increase of orientation angle, and increased with the pressure from 1 bar to 10 bar. Under atmospheric pressure, the CHF decreased with increasing the width of the heating surface, but there was a width value that could eliminate the influence of width. In addition, the CHF seemed to increase with the increase of thermal effusivity. However, there are still some inexplicable phenomena that need to be revealed by further research. Overall, this method is expected to be widely used in predicting the CHF of pool boiling on downward facing surfaces.

Experimental Research for CHF Sensitivity of Heat Flux Distribution under IVR Conditions

In-vessel retention (IVR) through external reactor vessel cooling (ERVC) is one of the most effective severe accident mitigation measures in the nuclear power plants. The most influential issues on the IVR strategy are in-vessel core melt evolution, the heat fluxes imposed on the lower head, and the external cooling of reactor pressurized vessel (RPV). In the molten pool research, there are mainly two different molten pool configurations: two layers and three layers. Based on the different distributions of heat flux in molten pool configurations, a new problem was raised: whether the in-vessel heat flux distribution will affect the CHF on the outer wall of RPV and further affect the effectiveness of IVR measures? A full-height external reactor vessel cooling and natural circulating facility was conducted to study the CHF sensitivity of different heat flux distributions. The experimental results show that the characteristics of natural circulation are similar and the CHF of the RPV lower head external surface is not obviously affected under the different heat flux distributions. The varying heat flux distribution during severe accident process will not threaten significantly the success of IVR strategy.

Cooling time and CHF enhancement of reactor vessel cooling process by flame sprayed porous coating in different orientations

External reactor vessel cooling (ERVC) is an effective strategy to achieve in-vessel retention (IVR) under severe nuclear plant accidents. Earlier transition from film boiling to nucleate boiling can lead to faster cooling and the critical heat flux (CHF) of the reactor pressure vessel (RPV) determines the success of IVR-ERVC. In this study, a porous coating of 316L was prepared on the surface of SA508 steel by flame spraying. Quenching experiments of plain surface and coating surface were performed to investigate the enhancement of reactor vessel cooling process at atmospheric pressure in saturated deionized water by using a variable orientation quenching device. The orientations of all experimental surfaces were varied between 0°(downward) and 90°(vertical). The experimental results show that the cooling time of coating surface was reduced by more than 56% compared with plain surface, which is mainly due to the reduction of duration in film boiling phase and the early appearance of Leidenfrost point (LFP). The CHF of coating surface achieved about 30% enhancement compared with plain surface in all orientations. In addition, the cooling time decreased and CHF increased with the increase of orientation because the bubbles separation was easier in the larger orientation angle. The visualization of cooling process was conducted in the vertical orientation and demonstrated that there were more nucleation sites of bubbles and the bubbles grew and separated faster in coating surface.

Research on Optimized Design of In-Vessel Retention–External Reactor Vessel Cooling Strategy and Negative Effect Assessment

Due to large uncertainty, the effectiveness of in-vessel retention (IVR) with external reactor vessel cooling (ERVC) for high-power reactors cannot be fully demonstrated during the transient process. To optimize the current IVR-ERVC strategy, the concept of IVR-ERVC with in-vessel injection (IVI) is studied. Two feasible IVI designs are proposed: 1) using the passive IVR water tank to implement simultaneous water injection in-vessel and ex-vessel and 2) using the passive severe accident dedicated in-vessel injection tanks (SADITs) to implement water injection in-vessel. The research on the feasibility and effectiveness of IVI is performed, and the corresponding negative effects are analyzed. The calculation results by using MIDAC show that the two proposed IVI concepts can greatly delay the accident progression in the core and reduce the decay heat peak of the molten pool in the lower head, thereby improving the effectiveness of the current IVR-ERVC strategy. And the negative effects are within acceptable ranges.

Heat transfer characteristics of single-layer and two-layer corium pool in elliptical lower head

In this study, we investigated the characteristics of natural convection heat transfer of single-layer and two-layer corium pool in the elliptical lower head by the numerical method. Numerical simulations were performed using FLUENT software with the WMLES (Wall-Modeled Large Eddy Simulation) turbulence model, solid-liquid phase change model and the VOF (Volume of Fluid) model. The heat transfer characteristics such as pool temperature field, flow velocity field and heat flux distribution were obtained. For the single-layer corium pool, different degrees of melting occurred on the wall surface from about 1712 mm arc length inside the inner wall to the liquid level. The transient velocity field of single-layer corium pool showed that the natural convection first occurs near the wall surface and then gradually forms in the center of the corium pool as the temperature rises. When the quasi-steady state is reached, the upper part of the pool forms strong turbulence, while the lower part has lower velocity and forms transverse flow. For the two-layer corium pool, the lower head wall is more severely melted than the single-layer configuration. Whereas, the crust at the wall near the stratified interface is only slightly melted due to the lower flow rate and weaker heat transfer capacity. The research revealed the flow and heat transfer mechanisms of corium pool, and provided reference for the technology design, system optimization and safety evaluation of the IVR (In-Vessel Retention) – ERVC (External Reactor Vessel Cooling) severe accident mitigation strategy for the large-scale advanced pressurized water reactor.