Muritala A Amidu | Khalifa University (original) (raw)

Papers by Muritala A Amidu

Research paper thumbnail of Influence of corium temperature, concrete composition and water injection time on concrete ablation during MCCI: New insights

Progress in Nuclear Energy, Feb 1, 2022

Molten corium, a mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural elemen... more Molten corium, a mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural elements, can originate in a nuclear plant accident after a reactor core meltdown. This un-cooled corium could penetrate through the reactor pressure vessel and cause concrete ablation via basement melt-through, a process known as Molten Corium Concrete Interaction (MCCI). The MCCI analysis because of its complex nature is still uncertain and needs thorough investigation of various parameters. In this study the use of CORQUENCH simulator is presented to model the molten corium, composition of concrete and heat transfer along with related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH is successfully utilized enabling the modeling of interaction between molten corium and concrete. The developed model is validated against experimental data at PWR and BWR conditions. The results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete was found to be the main controlling factor to mitigate ablation. An alternative to concrete is to utilize igneous rock (pyrolite) and this approach could lead to comparatively very low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a cooling agent) into the reactor cavity should also be optimized to enhance corium quenching to avoid ablation via basement melt-through. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus and ablation temperatures, respectively.

Research paper thumbnail of Experimental Study of Boiling Heat Transfer of Inclined Down-ward Facing Heated Curved Wall under Low Flow and Pressure Conditions

Applied Thermal Engineering

Research paper thumbnail of Sensitivity Analysis of Ex-Vessel Corium Coolability Models in MAAP5 Code for the Prediction of Molten Corium–Concrete Interaction after a Severe Accident Scenario

Energies

A postulated progressing severe accident scenario has been simulated using MAAP5 code with the fo... more A postulated progressing severe accident scenario has been simulated using MAAP5 code with the focus on ex-vessel cooling of molten corium in the reactor cavity. Various parameters associated with the prediction of molten corium–concrete interaction (MCCI) are identified. Accordingly, a sensitivity analysis is performed to assess the impact of these parameters on the predicted cavity floor erosion depth during this MCCI postulated accident. The sensitivity index of each variable parameter is determined using the Cotter indices method and Sobol′ indices method. At the early stage of the accident, the predicted cavity floor erosion depth is found to be highly sensitive to the downward heat transfer coefficient parameter with Cotter and Sobol′ indices of 94% and 50%, respectively. At the late phase of the accident, however, the cavity floor erosion depth becomes sensitive to melt eruption (Cotter index of 40%), water ingression (Cotter index of 13%), and particulate bed (Cotter index o...

Research paper thumbnail of Severe accident in high-power light water reactors: Mitigating strategies, assessment methods and research opportunities

Progress in Nuclear Energy, 2022

Research paper thumbnail of A New Insight Into Molten Corium Concrete Interaction With Concrete Ablation Analysis for Mitigation Scheme

Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, 2021

The study presents the use of CORQUENCH simulator to simultaneously model the molten corium, comp... more The study presents the use of CORQUENCH simulator to simultaneously model the molten corium, composition of concrete, molten corium heat transfer models and solve the related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH were successfully utilized which enabled the modeling of interaction between molten corium and concrete. The developed model was validated against experimental data and the results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating depth of ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete is the main controlling factor to mitigate ablation in the investigated case study. An alternative to concrete, could be a certain igneous rock (tested in this study), can lead to comparatively low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a...

Research paper thumbnail of A Critical Assessment of Nanoparticles Enhanced Phase Change Materials (Nepcms) for Latent Heat Energy Storage Applications

SSRN Electronic Journal, 2022

Research paper thumbnail of Numerical prediction of slug flow boiling heat transfer in the core-catcher cooling channel for severe accident mitigation in nuclear power plant

Nuclear Engineering and Design

This paper presents the steps followed to implement and validate a hybrid multiphase flow model i... more This paper presents the steps followed to implement and validate a hybrid multiphase flow model in the open-source code, OpenFOAM. The modeling approach couples the interface capturing model with the dispersed flow model. The resulting multiphase model can be used to predict the slug flow boiling regime. The flow regime in question occurs during the external cooling of a core-catcher and in-vessel retention (IVR) which are severe accident mitigation strategies. A distinctive key feature of this multiphase-type flow is the coexistence of large-scale slug vapor bubbles with both dispersed vapor bubbles and the carrying liquid phase. The slug vapor bubbles are generated from the coalescence of the smaller dispersed bubbles. Also, due to the tilted orientation of the core-catcher and reactor vessel lower head (for the IVR option), these large-scale bubbles remain in the vicinity of the heated surface, while being transported by the flow. This is due to the buoyancy force acting upward in these two design configurations. The latter phenomenon engenders the fact that a liquid film is occupying a thin layer separating the large bubbles from the heated surface. Under such flow conditions, the existing wall boiling model, commonly known as the (Rensselaer Polytechnic Institute) RPI model, has been demonstrated to be inadequate for the determination of the boiling heat transfer characteristics. Therefore, an extended near-wall boiling model accounting for the conduction heat flux across the liquid film (trapped underneath the slug bubbles) is formulated and implemented in this study. Using this enhanced model, the simulation of a slug flow boiling on a downward-facing heated surface produces a better prediction of the wall superheat than the original model. In addition, the morphologies of the vapor slug coexisting with dispersed bubbles are adequately captured and compared fairly well with experimental visualizations. This new multiphase model is then used to simulate a prototypical core-catcher cooling channel. Once again, a fair representation of the wall heat transfer is predicted in good agreement with measurements. Finally, it has been also successfully proven that under subcooled nucleate flow boiling conditions, the present model can reproduce the RPI model predictions.

Research paper thumbnail of Numerical prediction of slug flow boiling heat transfer in the core-catcher cooling channel for severe accident mitigation in nuclear power plant

Nuclear Engineering and Design, 2022

This paper presents the steps followed to implement and validate a hybrid multiphase flow model i... more This paper presents the steps followed to implement and validate a hybrid multiphase flow model in the open-source code, OpenFOAM. The modeling approach couples the interface capturing model with the dispersed flow model. The resulting multiphase model can be used to predict the slug flow boiling regime. The flow regime in question occurs during the external cooling of a core-catcher and in-vessel retention (IVR) which are severe accident mitigation strategies. A distinctive key feature of this multiphase-type flow is the coexistence of large-scale slug vapor bubbles with both dispersed vapor bubbles and the carrying liquid phase. The slug vapor bubbles are generated from the coalescence of the smaller dispersed bubbles. Also, due to the tilted orientation of the core-catcher and reactor vessel lower head (for the IVR option), these large-scale bubbles remain in the vicinity of the heated surface, while being transported by the flow. This is due to the buoyancy force acting upward in these two design configurations. The latter phenomenon engenders the fact that a liquid film is occupying a thin layer separating the large bubbles from the heated surface. Under such flow conditions, the existing wall boiling model, commonly known as the (Rensselaer Polytechnic Institute) RPI model, has been demonstrated to be inadequate for the determination of the boiling heat transfer characteristics. Therefore, an extended near-wall boiling model accounting for the conduction heat flux across the liquid film (trapped underneath the slug bubbles) is formulated and implemented in this study. Using this enhanced model, the simulation of a slug flow boiling on a downward-facing heated surface produces a better prediction of the wall superheat than the original model. In addition, the morphologies of the vapor slug coexisting with dispersed bubbles are adequately captured and compared fairly well with experimental visualizations. This new multiphase model is then used to simulate a prototypical core-catcher cooling channel. Once again, a fair representation of the wall heat transfer is predicted in good agreement with measurements. Finally, it has been also successfully proven that under subcooled nucleate flow boiling conditions, the present model can reproduce the RPI model predictions.

Research paper thumbnail of Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

Nuclear Engineering and Design

Abstract In the event of a core meltdown in a high-power reactor, the integrity of the reactor pr... more Abstract In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the ex-vessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

Research paper thumbnail of Toward Mechanistic Wall Heat Flux Partitioning Model for Fully Developed Nucleate Boiling

Journal of Heat Transfer

Mechanistic models developed to predict partial nucleate boiling are not adequate for fully devel... more Mechanistic models developed to predict partial nucleate boiling are not adequate for fully developed nucleate boiling due to differences in the prevailing heat transfer governing mechanisms. In place of the mechanistic model, several empirical correlations and semimechanistic models have been proposed over the years for the prediction of fully developed nucleate boiling as presented in this study but they are unsuitable for use in computational fluid dynamics (CFD) code. Recently, the simulation of fully developed nucleate boiling has become much more practical because of advancement in a computational method that involves the coupling of the interface capturing method (for slug bubbles) with the Eulerian multifluid model (for dispersed spherical bubbles). Nonetheless, there is a need for a mechanistic closure law for the fully developed nucleate boiling phenomenon that would complement this advancement in CFD. Toward this end, a mechanistic wall heat flux partitioning model for fu...

Research paper thumbnail of Safety assessment of AP1000: Common transients, analysis codes and research gaps

Nuclear Engineering and Design

Abstract The commercial operation of the AP1000 in China’s Sanmen nuclear power plant demonstrate... more Abstract The commercial operation of the AP1000 in China’s Sanmen nuclear power plant demonstrates the feasibility of reactors with advanced passive safety systems. However, being the first-of-a-kind, there is a need for robust, diverse, and independent safety assessment of the plant’s components and systems. This study presents a critical review of AP1000 transients, categorizes and evaluates current safety assessment codes, and discusses their application to safety analysis of AP1000. We extensively discuss the AP1000 safety assessment methods and enumerate sources of uncertainties in codes that have been used to assess a wide spectrum of AP1000 transients. In addition, we identified basic developmental issues in different system codes and crucial problems with their applications to AP1000. Furthermore, we give insights into optimized simulation techniques and advanced modeling approaches for high fidelity computation. As unique contributions, key issues such as the reliability of the passive safety systems, scaling, verification and validation experiments necessary to enhance the safety of AP1000 are discussed. The significant research gaps, future research direction, and current safety issues presented in this work also serve as an important body of knowledge towards a safe and reliable operation of future AP1000 fleets.

Research paper thumbnail of A hybrid multiphase flow model for the prediction of both low and high void fraction nucleate boiling regimes

Applied Thermal Engineering

Abstract The improvement of a hybrid (a combination of Volume-of-Fluid (VOF) and Eulerian model) ... more Abstract The improvement of a hybrid (a combination of Volume-of-Fluid (VOF) and Eulerian model) multiphase flow solver for the numerical prediction of high- and low-void fraction flow boiling regimes is presented in this article. These flow regimes could simultaneously occur in core-catcher and in-vessel retention-external reactor cooling systems during a severe accident in nuclear power plant. To enhance the prediction of the low void fraction boiling regime, a turbulent rough wall function model is implemented in the hybrid model to reproduce the impact of coarseness induced by the existence of growing bubbles along the heating wall on the liquid velocity profile. With this wall function, a more accurate prediction of the radial velocity profile is achieved within the uncertainty of the velocity measurements. Moreover, an improved prediction of the radial void fraction is achieved using the model proposed by Lopez de Bertodano for turbulence dispersion force without compromising the prediction of the radial gas velocity profile and radial liquid temperature profile. Although the hybrid model shows potential in capturing the interface and dynamic behavior of large-scale bubbles (vapor slug) for high void fraction regime, the predicted wall superheat is higher than the measured values. This highlighted the need for the extension of the present wall boiling model to cover flow boiling involving sliding vapor slugs on the heated wall.

Research paper thumbnail of Bubble-induced enhancement of single-phase liquid forced convection heat transfer during subcooled nucleate flow boiling

Annals of Nuclear Energy

Abstract In two-phase flow boiling systems, wall heat flux partitioning model is used in computat... more Abstract In two-phase flow boiling systems, wall heat flux partitioning model is used in computational fluid dynamics (CFD) codes to quantify the heat transfer to liquid and heat transfer due to vapor generation separately. Recently, component-wise validation of each of the components (convection, quenching, and evaporation) of the wall heat flux-partitioning model was performed and it was observed that the bubble-induced enhancement of the single-phase convection heat transfer coefficient could not be adequately captured in the existing model. Therefore, leveraging these recent direct experimental measurements of the components of the wall heat flux during subcooled flow boiling, a simple model that depends on both the size and population density of the bubble is proposed for hydrodynamic roughness. This model adequately accounts for the heat transfer enhancement due to the presence of bubbles on the heater surface and it shows a better prediction of the experimental data than the existing model.

Research paper thumbnail of Modeling and simulation of flow boiling heat transfer on a downward-facing heating wall in the presence of vapor slugs

Nuclear Engineering and Design

Abstract Nucleate boiling on downward facing heating walls such as those found in IVR-ERVC and co... more Abstract Nucleate boiling on downward facing heating walls such as those found in IVR-ERVC and core catcher systems of advanced LWRs is characterized by bubbles of markedly different scales (small bubbles and vapor slugs). The heat transfer governing mechanism of deformable vapor slugs is different from that of small spherical bubbles. Therefore, the two-fluid wall-boiling model developed based on the characteristic of small-scale nucleate bubbles might not be sufficient for prediction of wall boiling heat transfer on a downward facing heating surface. In this article, a hybrid wall-boiling model is formulated and implemented in the solution framework of the Eulerian-Eulerian model coupled with a large-scale interface model and adaptive interface sharpening scheme. This approach gives a realistic representation of the physical phenomena present in flow boiling on a downward facing heating wall, and the validation against experimental data showed similar performance to the two-fluid model in terms of the prediction of wall boiling heat transfer.

Research paper thumbnail of Semi-mechanistic model for the interfacial velocity of gravity-driven laminar wavy film flow and its validation using infrared particle tracing velocimetry

Heat and Mass Transfer

Interfacial velocity is an important parameter in the modeling of momentum transfer for predictio... more Interfacial velocity is an important parameter in the modeling of momentum transfer for prediction of heat- and mass-transfer during film-wise condensation. In this study, the interfacial velocity is modeled using an empirical power-law velocity profile with the assumption that the interfacial shear stress is negligible compared with the wall shear stress. A non-intrusive infrared particle tracking velocimetry (IR-PTV) measurement technique is used to validate a newly proposed semi-mechanistic model for the interfacial velocity of a gravity-driven laminar wavy film flow. The proposed model predicts measured interfacial velocities reasonably well and could serve as a closure relation in estimating the film-wise condensation heat transfer coefficient.

Research paper thumbnail of Direct experimental measurement for partitioning of wall heat flux during subcooled flow boiling: Effect of bubble areas of influence factor

International Journal of Heat and Mass Transfer

Abstract Heat transfer models in liquid-vapor two-phase flow with wall boiling rely on the wall h... more Abstract Heat transfer models in liquid-vapor two-phase flow with wall boiling rely on the wall heat flux partitioning to quantify heat transfer to liquid and vapor separately. Several wall heat flux partitioning models have been proposed over the years based on variety of heat transfer mechanisms, but the three basic mechanisms that form the core of these models are liquid convection, surface quenching and evaporation heat transfer. A key parameter commonly used to determine the relative contribution made by each mechanism is area fraction of influence of bubble which is determined by multiplying maximum bubble projected area fraction with bubble area of influence factor (K). In classic wall heat flux partitioning models, K accounts for the area within which heat is transferred to liquid that moves in towards the heated wall as bubbles lift-off. The value of K has been a subject of controversy over the years with no unanimous conclusion among researchers. Therefore, in this paper, advanced diagnostic approach involving the combination of infrared thermometry and total reflection principle was employed to experimentally study nucleate flow boiling. Rigorous data analyses was performed to partition the wall heat flux into the aforementioned three basic heat transfer mechanisms using different values of K. All three heat transfer mechanisms were significantly sensitive to varying values of K, but setting K = 0.5 with percentage uncertainties of −60%/+50% closely predicted the experimental measurements. In addition, overlapping area of influence due to merging bubbles was observed to be significant in the model at high heat flux condition and must be discounted to get the true bubble area of influence. A correction method for the overlapping area of influence was therefore proposed to enhance accuracy of the predictive model.

Research paper thumbnail of Performance analyses of a steam condensation tube immersed in a saturated water pool: Effects of tube inclination

Nuclear Engineering and Design

Abstract Two-phase heat exchanger immersed in a passive containment cooling water storage tank (P... more Abstract Two-phase heat exchanger immersed in a passive containment cooling water storage tank (PCCWST) is a key component of the passive containment cooling system (PCCS) in light water reactors. Condensation and boiling heat transfer phenomena taking place on the inside and outside walls are crucial to the performance of the heat exchanger tube. The performances of such phenomena can be readily affected by the inclination angle of the heat exchanger tube under buoyancy-driven convection conditions in a PCCWST. However, to date no systematic examination of the effects of inclination angle on pool heat exchanger performance has been reported. This paper presents the results of numerical and experimental analyses of how inclination angle affects the performance of a single steam condensation tube immersed in a saturated water pool. To concurrently predict the two-phase heat transfer processes inside and outside the heat exchanger tube, an explicit heat structure coupling of two thermal-hydraulic codes was implemented using open media models. An experimental facility was also constructed to test a single-tube heat exchanger under the same conditions as the simulation. A comparison of simulation data and experimental results obtained for the reference case (tube inclination of 30°) confirmed that the coupled code could predict the heat transfer rate in the pool heat exchanger within the error limits of the experimental measurements. In light of this, additional simulations and experiments were conducted at varying inclination angles, from 3° to 90°. The results of simulations and experimental studies revealed that the heat transfer rate of a heat exchanger tube in a saturated pool is hardly influenced by tube inclination. Although an increase in inclination angle caused the vapor slug to spread around the heat exchanger tube surface, preventing liquid from getting to the surface of the tube, thereby decreasing boiling heat transfer coefficients, this was compensated for by an increase in condensation heat transfer coefficients as the inclination angle increased as a result of accelerated condensate liquid film in the tube. Thus, the overall effect of inclination angle on the thermal performance of a single-tube pool heat exchanger is inconsequential.

Research paper thumbnail of Numerical investigation of nanoparticles slip mechanisms impact on the natural convection heat transfer characteristics of nanofluids in an enclosure

Scientific Reports

This study intends to give qualitative results toward the understanding of different slip mechani... more This study intends to give qualitative results toward the understanding of different slip mechanisms impact on the natural heat transfer performance of nanofluids. The slip mechanisms considered in this study are Brownian diffusion, thermophoretic diffusion, and sedimentation. This study compares three different Eulerian nanofluid models; Single-phase, two-phase, and a third model that consists of incorporating the three slip mechanisms in a two-phase drift-flux. These slip mechanisms are found to have different impacts depending on the nanoparticle concentration, where this effect ranges from negligible to dominant. It has been reported experimentally in the literature that, with high nanoparticle volume fraction the heat transfer deteriorates. Admittingly, classical nanofluid models are known to underpredict this impairment. To address this discrepancy, this study focuses on the effect of thermophoretic diffusion and sedimentation outcome as these two mechanisms turn out to be inf...

Research paper thumbnail of Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

Nuclear Engineering and Design

In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure ve... more In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the exvessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

Research paper thumbnail of A hybrid multiphase flow model for the prediction of both low and high void fraction nucleate boiling regimes

Applied Thermal Engineering

• A hybrid VOF-Eulerian multifluid model is presented. • A turbulent rough wall function is propo... more • A hybrid VOF-Eulerian multifluid model is presented. • A turbulent rough wall function is proposed to enhance the prediction of the velocity profile. • The hybrid model reasonably predict the features of low void fraction flow boiling. • The hybrid model could captures large scale interface in high void fraction flow boiling. • Need to extend the wall boiling model to accommodate heat transfer beneath vapor slugs.

Research paper thumbnail of Influence of corium temperature, concrete composition and water injection time on concrete ablation during MCCI: New insights

Progress in Nuclear Energy, Feb 1, 2022

Molten corium, a mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural elemen... more Molten corium, a mixture of molten nuclear fuel, cladding, thermo-hydraulic and structural elements, can originate in a nuclear plant accident after a reactor core meltdown. This un-cooled corium could penetrate through the reactor pressure vessel and cause concrete ablation via basement melt-through, a process known as Molten Corium Concrete Interaction (MCCI). The MCCI analysis because of its complex nature is still uncertain and needs thorough investigation of various parameters. In this study the use of CORQUENCH simulator is presented to model the molten corium, composition of concrete and heat transfer along with related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH is successfully utilized enabling the modeling of interaction between molten corium and concrete. The developed model is validated against experimental data at PWR and BWR conditions. The results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete was found to be the main controlling factor to mitigate ablation. An alternative to concrete is to utilize igneous rock (pyrolite) and this approach could lead to comparatively very low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a cooling agent) into the reactor cavity should also be optimized to enhance corium quenching to avoid ablation via basement melt-through. The concrete ablation mechanisms during MCCI are very case-dependent on the concrete solidus, liquidus and ablation temperatures, respectively.

Research paper thumbnail of Experimental Study of Boiling Heat Transfer of Inclined Down-ward Facing Heated Curved Wall under Low Flow and Pressure Conditions

Applied Thermal Engineering

Research paper thumbnail of Sensitivity Analysis of Ex-Vessel Corium Coolability Models in MAAP5 Code for the Prediction of Molten Corium–Concrete Interaction after a Severe Accident Scenario

Energies

A postulated progressing severe accident scenario has been simulated using MAAP5 code with the fo... more A postulated progressing severe accident scenario has been simulated using MAAP5 code with the focus on ex-vessel cooling of molten corium in the reactor cavity. Various parameters associated with the prediction of molten corium–concrete interaction (MCCI) are identified. Accordingly, a sensitivity analysis is performed to assess the impact of these parameters on the predicted cavity floor erosion depth during this MCCI postulated accident. The sensitivity index of each variable parameter is determined using the Cotter indices method and Sobol′ indices method. At the early stage of the accident, the predicted cavity floor erosion depth is found to be highly sensitive to the downward heat transfer coefficient parameter with Cotter and Sobol′ indices of 94% and 50%, respectively. At the late phase of the accident, however, the cavity floor erosion depth becomes sensitive to melt eruption (Cotter index of 40%), water ingression (Cotter index of 13%), and particulate bed (Cotter index o...

Research paper thumbnail of Severe accident in high-power light water reactors: Mitigating strategies, assessment methods and research opportunities

Progress in Nuclear Energy, 2022

Research paper thumbnail of A New Insight Into Molten Corium Concrete Interaction With Concrete Ablation Analysis for Mitigation Scheme

Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, 2021

The study presents the use of CORQUENCH simulator to simultaneously model the molten corium, comp... more The study presents the use of CORQUENCH simulator to simultaneously model the molten corium, composition of concrete, molten corium heat transfer models and solve the related chemical reactions. Using this modeling technique, the chemical reaction capabilities of CORQUENCH were successfully utilized which enabled the modeling of interaction between molten corium and concrete. The developed model was validated against experimental data and the results showed that the temperature of corium, composition of concrete and water injection time have a pronounced effect on mitigating depth of ablation and reactor integrity in case of a nuclear accident. In addition, the composition of concrete is the main controlling factor to mitigate ablation in the investigated case study. An alternative to concrete, could be a certain igneous rock (tested in this study), can lead to comparatively low rates of ablation due to its high thermal resistant properties. Furthermore, the injection of water (as a...

Research paper thumbnail of A Critical Assessment of Nanoparticles Enhanced Phase Change Materials (Nepcms) for Latent Heat Energy Storage Applications

SSRN Electronic Journal, 2022

Research paper thumbnail of Numerical prediction of slug flow boiling heat transfer in the core-catcher cooling channel for severe accident mitigation in nuclear power plant

Nuclear Engineering and Design

This paper presents the steps followed to implement and validate a hybrid multiphase flow model i... more This paper presents the steps followed to implement and validate a hybrid multiphase flow model in the open-source code, OpenFOAM. The modeling approach couples the interface capturing model with the dispersed flow model. The resulting multiphase model can be used to predict the slug flow boiling regime. The flow regime in question occurs during the external cooling of a core-catcher and in-vessel retention (IVR) which are severe accident mitigation strategies. A distinctive key feature of this multiphase-type flow is the coexistence of large-scale slug vapor bubbles with both dispersed vapor bubbles and the carrying liquid phase. The slug vapor bubbles are generated from the coalescence of the smaller dispersed bubbles. Also, due to the tilted orientation of the core-catcher and reactor vessel lower head (for the IVR option), these large-scale bubbles remain in the vicinity of the heated surface, while being transported by the flow. This is due to the buoyancy force acting upward in these two design configurations. The latter phenomenon engenders the fact that a liquid film is occupying a thin layer separating the large bubbles from the heated surface. Under such flow conditions, the existing wall boiling model, commonly known as the (Rensselaer Polytechnic Institute) RPI model, has been demonstrated to be inadequate for the determination of the boiling heat transfer characteristics. Therefore, an extended near-wall boiling model accounting for the conduction heat flux across the liquid film (trapped underneath the slug bubbles) is formulated and implemented in this study. Using this enhanced model, the simulation of a slug flow boiling on a downward-facing heated surface produces a better prediction of the wall superheat than the original model. In addition, the morphologies of the vapor slug coexisting with dispersed bubbles are adequately captured and compared fairly well with experimental visualizations. This new multiphase model is then used to simulate a prototypical core-catcher cooling channel. Once again, a fair representation of the wall heat transfer is predicted in good agreement with measurements. Finally, it has been also successfully proven that under subcooled nucleate flow boiling conditions, the present model can reproduce the RPI model predictions.

Research paper thumbnail of Numerical prediction of slug flow boiling heat transfer in the core-catcher cooling channel for severe accident mitigation in nuclear power plant

Nuclear Engineering and Design, 2022

This paper presents the steps followed to implement and validate a hybrid multiphase flow model i... more This paper presents the steps followed to implement and validate a hybrid multiphase flow model in the open-source code, OpenFOAM. The modeling approach couples the interface capturing model with the dispersed flow model. The resulting multiphase model can be used to predict the slug flow boiling regime. The flow regime in question occurs during the external cooling of a core-catcher and in-vessel retention (IVR) which are severe accident mitigation strategies. A distinctive key feature of this multiphase-type flow is the coexistence of large-scale slug vapor bubbles with both dispersed vapor bubbles and the carrying liquid phase. The slug vapor bubbles are generated from the coalescence of the smaller dispersed bubbles. Also, due to the tilted orientation of the core-catcher and reactor vessel lower head (for the IVR option), these large-scale bubbles remain in the vicinity of the heated surface, while being transported by the flow. This is due to the buoyancy force acting upward in these two design configurations. The latter phenomenon engenders the fact that a liquid film is occupying a thin layer separating the large bubbles from the heated surface. Under such flow conditions, the existing wall boiling model, commonly known as the (Rensselaer Polytechnic Institute) RPI model, has been demonstrated to be inadequate for the determination of the boiling heat transfer characteristics. Therefore, an extended near-wall boiling model accounting for the conduction heat flux across the liquid film (trapped underneath the slug bubbles) is formulated and implemented in this study. Using this enhanced model, the simulation of a slug flow boiling on a downward-facing heated surface produces a better prediction of the wall superheat than the original model. In addition, the morphologies of the vapor slug coexisting with dispersed bubbles are adequately captured and compared fairly well with experimental visualizations. This new multiphase model is then used to simulate a prototypical core-catcher cooling channel. Once again, a fair representation of the wall heat transfer is predicted in good agreement with measurements. Finally, it has been also successfully proven that under subcooled nucleate flow boiling conditions, the present model can reproduce the RPI model predictions.

Research paper thumbnail of Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

Nuclear Engineering and Design

Abstract In the event of a core meltdown in a high-power reactor, the integrity of the reactor pr... more Abstract In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the ex-vessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

Research paper thumbnail of Toward Mechanistic Wall Heat Flux Partitioning Model for Fully Developed Nucleate Boiling

Journal of Heat Transfer

Mechanistic models developed to predict partial nucleate boiling are not adequate for fully devel... more Mechanistic models developed to predict partial nucleate boiling are not adequate for fully developed nucleate boiling due to differences in the prevailing heat transfer governing mechanisms. In place of the mechanistic model, several empirical correlations and semimechanistic models have been proposed over the years for the prediction of fully developed nucleate boiling as presented in this study but they are unsuitable for use in computational fluid dynamics (CFD) code. Recently, the simulation of fully developed nucleate boiling has become much more practical because of advancement in a computational method that involves the coupling of the interface capturing method (for slug bubbles) with the Eulerian multifluid model (for dispersed spherical bubbles). Nonetheless, there is a need for a mechanistic closure law for the fully developed nucleate boiling phenomenon that would complement this advancement in CFD. Toward this end, a mechanistic wall heat flux partitioning model for fu...

Research paper thumbnail of Safety assessment of AP1000: Common transients, analysis codes and research gaps

Nuclear Engineering and Design

Abstract The commercial operation of the AP1000 in China’s Sanmen nuclear power plant demonstrate... more Abstract The commercial operation of the AP1000 in China’s Sanmen nuclear power plant demonstrates the feasibility of reactors with advanced passive safety systems. However, being the first-of-a-kind, there is a need for robust, diverse, and independent safety assessment of the plant’s components and systems. This study presents a critical review of AP1000 transients, categorizes and evaluates current safety assessment codes, and discusses their application to safety analysis of AP1000. We extensively discuss the AP1000 safety assessment methods and enumerate sources of uncertainties in codes that have been used to assess a wide spectrum of AP1000 transients. In addition, we identified basic developmental issues in different system codes and crucial problems with their applications to AP1000. Furthermore, we give insights into optimized simulation techniques and advanced modeling approaches for high fidelity computation. As unique contributions, key issues such as the reliability of the passive safety systems, scaling, verification and validation experiments necessary to enhance the safety of AP1000 are discussed. The significant research gaps, future research direction, and current safety issues presented in this work also serve as an important body of knowledge towards a safe and reliable operation of future AP1000 fleets.

Research paper thumbnail of A hybrid multiphase flow model for the prediction of both low and high void fraction nucleate boiling regimes

Applied Thermal Engineering

Abstract The improvement of a hybrid (a combination of Volume-of-Fluid (VOF) and Eulerian model) ... more Abstract The improvement of a hybrid (a combination of Volume-of-Fluid (VOF) and Eulerian model) multiphase flow solver for the numerical prediction of high- and low-void fraction flow boiling regimes is presented in this article. These flow regimes could simultaneously occur in core-catcher and in-vessel retention-external reactor cooling systems during a severe accident in nuclear power plant. To enhance the prediction of the low void fraction boiling regime, a turbulent rough wall function model is implemented in the hybrid model to reproduce the impact of coarseness induced by the existence of growing bubbles along the heating wall on the liquid velocity profile. With this wall function, a more accurate prediction of the radial velocity profile is achieved within the uncertainty of the velocity measurements. Moreover, an improved prediction of the radial void fraction is achieved using the model proposed by Lopez de Bertodano for turbulence dispersion force without compromising the prediction of the radial gas velocity profile and radial liquid temperature profile. Although the hybrid model shows potential in capturing the interface and dynamic behavior of large-scale bubbles (vapor slug) for high void fraction regime, the predicted wall superheat is higher than the measured values. This highlighted the need for the extension of the present wall boiling model to cover flow boiling involving sliding vapor slugs on the heated wall.

Research paper thumbnail of Bubble-induced enhancement of single-phase liquid forced convection heat transfer during subcooled nucleate flow boiling

Annals of Nuclear Energy

Abstract In two-phase flow boiling systems, wall heat flux partitioning model is used in computat... more Abstract In two-phase flow boiling systems, wall heat flux partitioning model is used in computational fluid dynamics (CFD) codes to quantify the heat transfer to liquid and heat transfer due to vapor generation separately. Recently, component-wise validation of each of the components (convection, quenching, and evaporation) of the wall heat flux-partitioning model was performed and it was observed that the bubble-induced enhancement of the single-phase convection heat transfer coefficient could not be adequately captured in the existing model. Therefore, leveraging these recent direct experimental measurements of the components of the wall heat flux during subcooled flow boiling, a simple model that depends on both the size and population density of the bubble is proposed for hydrodynamic roughness. This model adequately accounts for the heat transfer enhancement due to the presence of bubbles on the heater surface and it shows a better prediction of the experimental data than the existing model.

Research paper thumbnail of Modeling and simulation of flow boiling heat transfer on a downward-facing heating wall in the presence of vapor slugs

Nuclear Engineering and Design

Abstract Nucleate boiling on downward facing heating walls such as those found in IVR-ERVC and co... more Abstract Nucleate boiling on downward facing heating walls such as those found in IVR-ERVC and core catcher systems of advanced LWRs is characterized by bubbles of markedly different scales (small bubbles and vapor slugs). The heat transfer governing mechanism of deformable vapor slugs is different from that of small spherical bubbles. Therefore, the two-fluid wall-boiling model developed based on the characteristic of small-scale nucleate bubbles might not be sufficient for prediction of wall boiling heat transfer on a downward facing heating surface. In this article, a hybrid wall-boiling model is formulated and implemented in the solution framework of the Eulerian-Eulerian model coupled with a large-scale interface model and adaptive interface sharpening scheme. This approach gives a realistic representation of the physical phenomena present in flow boiling on a downward facing heating wall, and the validation against experimental data showed similar performance to the two-fluid model in terms of the prediction of wall boiling heat transfer.

Research paper thumbnail of Semi-mechanistic model for the interfacial velocity of gravity-driven laminar wavy film flow and its validation using infrared particle tracing velocimetry

Heat and Mass Transfer

Interfacial velocity is an important parameter in the modeling of momentum transfer for predictio... more Interfacial velocity is an important parameter in the modeling of momentum transfer for prediction of heat- and mass-transfer during film-wise condensation. In this study, the interfacial velocity is modeled using an empirical power-law velocity profile with the assumption that the interfacial shear stress is negligible compared with the wall shear stress. A non-intrusive infrared particle tracking velocimetry (IR-PTV) measurement technique is used to validate a newly proposed semi-mechanistic model for the interfacial velocity of a gravity-driven laminar wavy film flow. The proposed model predicts measured interfacial velocities reasonably well and could serve as a closure relation in estimating the film-wise condensation heat transfer coefficient.

Research paper thumbnail of Direct experimental measurement for partitioning of wall heat flux during subcooled flow boiling: Effect of bubble areas of influence factor

International Journal of Heat and Mass Transfer

Abstract Heat transfer models in liquid-vapor two-phase flow with wall boiling rely on the wall h... more Abstract Heat transfer models in liquid-vapor two-phase flow with wall boiling rely on the wall heat flux partitioning to quantify heat transfer to liquid and vapor separately. Several wall heat flux partitioning models have been proposed over the years based on variety of heat transfer mechanisms, but the three basic mechanisms that form the core of these models are liquid convection, surface quenching and evaporation heat transfer. A key parameter commonly used to determine the relative contribution made by each mechanism is area fraction of influence of bubble which is determined by multiplying maximum bubble projected area fraction with bubble area of influence factor (K). In classic wall heat flux partitioning models, K accounts for the area within which heat is transferred to liquid that moves in towards the heated wall as bubbles lift-off. The value of K has been a subject of controversy over the years with no unanimous conclusion among researchers. Therefore, in this paper, advanced diagnostic approach involving the combination of infrared thermometry and total reflection principle was employed to experimentally study nucleate flow boiling. Rigorous data analyses was performed to partition the wall heat flux into the aforementioned three basic heat transfer mechanisms using different values of K. All three heat transfer mechanisms were significantly sensitive to varying values of K, but setting K = 0.5 with percentage uncertainties of −60%/+50% closely predicted the experimental measurements. In addition, overlapping area of influence due to merging bubbles was observed to be significant in the model at high heat flux condition and must be discounted to get the true bubble area of influence. A correction method for the overlapping area of influence was therefore proposed to enhance accuracy of the predictive model.

Research paper thumbnail of Performance analyses of a steam condensation tube immersed in a saturated water pool: Effects of tube inclination

Nuclear Engineering and Design

Abstract Two-phase heat exchanger immersed in a passive containment cooling water storage tank (P... more Abstract Two-phase heat exchanger immersed in a passive containment cooling water storage tank (PCCWST) is a key component of the passive containment cooling system (PCCS) in light water reactors. Condensation and boiling heat transfer phenomena taking place on the inside and outside walls are crucial to the performance of the heat exchanger tube. The performances of such phenomena can be readily affected by the inclination angle of the heat exchanger tube under buoyancy-driven convection conditions in a PCCWST. However, to date no systematic examination of the effects of inclination angle on pool heat exchanger performance has been reported. This paper presents the results of numerical and experimental analyses of how inclination angle affects the performance of a single steam condensation tube immersed in a saturated water pool. To concurrently predict the two-phase heat transfer processes inside and outside the heat exchanger tube, an explicit heat structure coupling of two thermal-hydraulic codes was implemented using open media models. An experimental facility was also constructed to test a single-tube heat exchanger under the same conditions as the simulation. A comparison of simulation data and experimental results obtained for the reference case (tube inclination of 30°) confirmed that the coupled code could predict the heat transfer rate in the pool heat exchanger within the error limits of the experimental measurements. In light of this, additional simulations and experiments were conducted at varying inclination angles, from 3° to 90°. The results of simulations and experimental studies revealed that the heat transfer rate of a heat exchanger tube in a saturated pool is hardly influenced by tube inclination. Although an increase in inclination angle caused the vapor slug to spread around the heat exchanger tube surface, preventing liquid from getting to the surface of the tube, thereby decreasing boiling heat transfer coefficients, this was compensated for by an increase in condensation heat transfer coefficients as the inclination angle increased as a result of accelerated condensate liquid film in the tube. Thus, the overall effect of inclination angle on the thermal performance of a single-tube pool heat exchanger is inconsequential.

Research paper thumbnail of Numerical investigation of nanoparticles slip mechanisms impact on the natural convection heat transfer characteristics of nanofluids in an enclosure

Scientific Reports

This study intends to give qualitative results toward the understanding of different slip mechani... more This study intends to give qualitative results toward the understanding of different slip mechanisms impact on the natural heat transfer performance of nanofluids. The slip mechanisms considered in this study are Brownian diffusion, thermophoretic diffusion, and sedimentation. This study compares three different Eulerian nanofluid models; Single-phase, two-phase, and a third model that consists of incorporating the three slip mechanisms in a two-phase drift-flux. These slip mechanisms are found to have different impacts depending on the nanoparticle concentration, where this effect ranges from negligible to dominant. It has been reported experimentally in the literature that, with high nanoparticle volume fraction the heat transfer deteriorates. Admittingly, classical nanofluid models are known to underpredict this impairment. To address this discrepancy, this study focuses on the effect of thermophoretic diffusion and sedimentation outcome as these two mechanisms turn out to be inf...

Research paper thumbnail of Investigation of the pressure vessel lower head potential failure under IVR-ERVC condition during a severe accident scenario in APR1400 reactors

Nuclear Engineering and Design

In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure ve... more In the event of a core meltdown in a high-power reactor, the integrity of the reactor pressure vessel is presumably protected by severe accident mitigation systems such as in-vessel retention external reactor vessel cooling (IVR-ERVC). However, in the late phase of the accident, two possible locations on the RPV are prone to failure: the location of the focusing effect and location of in-core instrument penetration. These two potential points of damage in the RPV are investigated in this study. A numerical model for the prediction of the natural convection, melting, and solidification processes for IVR-ERVC is presented. The model is based on the enthalpy-porosity approach with an extension for continuous liquid fraction function. The model is implemented in open-source field operation and manipulation (OpenFOAM) computational fluid dynamic code to produce a new solver which is based on the combination of conjugate heat transfer solver and buoyant-driven natural convection solver and the new solver is validated against the melting Gallium experimental test, in-core instrumentation failure experimental test, and BALI experimental test. This numerical model is applied for the investigation of the RPV rupture at the location of the focusing effect and in-core instrumentation penetrations. Severe ablations of the cladding and the weld materials are observed at a heat load of about ~1800 K which is expected to lead to the ejection of the penetration tubes if the force holding the penetration tube in place is lower than the force exerted by the system pressure. Subsequently, a two-layer IVR configuration is assessed and the integrity of the RPV is found not to be compromised under external reactor vessel cooling. However, in the case of a boiling crisis, the temperature of the ex-vessel wall is expected to rise quickly and this is simulated by increasing the exvessel wall temperature. The RPV is found to fail near the beltline due to a phenomenon known as focusing effect when the ex-vessel wall temperature rises above 1200 K.

Research paper thumbnail of A hybrid multiphase flow model for the prediction of both low and high void fraction nucleate boiling regimes

Applied Thermal Engineering

• A hybrid VOF-Eulerian multifluid model is presented. • A turbulent rough wall function is propo... more • A hybrid VOF-Eulerian multifluid model is presented. • A turbulent rough wall function is proposed to enhance the prediction of the velocity profile. • The hybrid model reasonably predict the features of low void fraction flow boiling. • The hybrid model could captures large scale interface in high void fraction flow boiling. • Need to extend the wall boiling model to accommodate heat transfer beneath vapor slugs.