Nuclear Power Plant Containment Research Articles

Safety assessment of generation III nuclear power plant buildings subjected to commercial aircraft crash part III: Engine missile impacting SC plate

Abstract Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part III, the local damage of the rigid components of aircraft, e.g., engine and landing gear, impacting the steel concrete (SC) structures of NPP containment is mainly discussed. Two typical SC target panels with the thicknesses of 40 mm and 100 mm, as well as the steel cylindrical projectile with a mass of 2.15 kg and a diameter of 80 mm are fabricated. By using a large-caliber air gas gun, both the projectile penetration and perforation test are conducted, in which the striking velocities were ranged from 96 m/s to 157 m/s. The bulging velocity and the maximal deflection of rear steel plate, as well as penetration depth of projectile are derived, and the local deformation and failure modes of SC panels are assessed experimentally. Then, the commercial finite element program LS-DYNA is utilized to perform the numerical simulations, by comparisons with the experimental and simulated projectile impact process and SC panel damage, the numerical algorithm, constitutive models and the corresponding parameters are verified. The present work can provide helpful references for the evaluation of the local impact resistance of NPP buildings against the aircraft engine.

Safety assessment of Generation III nuclear power plant buildings subjected to commercial aircraft crash Part II: Structural damage and vibrations

Abstract Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part II, based on the verified finite element (FE) models of aircrafts Airbus A320 and A380, as well as the NPP containment and auxiliary buildings in Part I of this paper, the whole collision process is reproduced numerically by adopting the coupled missile-target interaction approach with the finite element code LS-DYNA. The impact induced damage of NPP plant under four impact locations of containment (cylinder, air intake, conical roof and PCS water tank) and two impact locations of auxiliary buildings (exterior wall and roof of spent fuel pool room) are evaluated. Furthermore, by considering the inner structures in the containment and raft foundation of NPP, the structural vibration analyses are conducted under two impact locations (middle height of cylinder, main control room in the auxiliary buildings). It indicates that, within the discussed scenarios, NPP structures can withstand the impact of both two aircrafts, while the functionality of internal equipment on higher floors will be affected to some extent under impact induced vibrations, and A380 aircraft will cause more serious structural damage and vibrations than A320 aircraft. The present work can provide helpful references to assess the safety of the structures and inner equipment of NPP plant under commercial aircraft impact.

Numerical study of the stratification erosion benchmark for NPPs containment using CFD code GASFLOW-MPI

During a severe accident involving a Nuclear Power Plant (NPP), a hydrogen-enriched stratification may occur in the dome of the containment. This is due to the gaseous density difference of a steam/hydrogen turbulent jet released from the break. This accident scenario is one of the high-ranking phenomena identified in the Phenomena Identification and Ranking Table (PIRT) for containment safety thermal hydraulics. In this work, the temporal evolution of the stratification erosion is analysed using the parallel CFD code GASFLOW-MPI with comparison to the experimental data from the THAI TH-20 benchmark, where helium is used as a hydrogen simulate. Identifying a suitable turbulence model is an important issue for a successful stratification erosion simulation due to turbulent mixing between the stratification and turbulent jet featured with the sharp velocity and density gradients. Three well-known turbulence models: 1. The classic k-ε model, 2. Large Eddy Simulation (LES) model and 3. Detached Eddy Simulation (DES) model are employed to evaluate their performances. Due to the negative buoyancy effect resulting from the light gas layer, the jet gradually loses its initial momentum and cannot penetrate the stratified layer in depth. The interaction Froude (Fr) number which is used to categorize the mixing mechanism is close to 1 in this work, indicating that the erosion process of the stratified layer is dominated by both the molecular diffusion and the momentum transport. The experimental data measured at different heights in THAI TH-20 benchmark, from the pure air region at the lower containment region up into the large helium concentration gradient region, and further up to high helium concentration region at the top of the containment. The calculated results show that the LES and DES turbulence models can capture time histories of helium concentration at all of the sensor points accurately, while the k-ε turbulence model predictions are always delayed due to its overestimating the turbulence mixing in the stratified layer. Moreover, compared with the k-ε model, more detailed turbulent eddy structure is resolved by the LES model and DES model, whose frequency satisfies the negative −5/3 energy decay rate for a wide region of frequency. Therefore, the LES and DES models are recommended for the stratification erosion simulation. This can provide guidance for nuclear engineers analysing hydrogen distribution phenomena during severe accidents.

ПРИМЕНЕНИЕ ЭКСПЕРТНОЙ СИСТЕМЫ ОЦЕНКИ НАПРЯЖЕННО-ДЕФОРМИРУЕМОГО СОСТОЯНИЯ ЗАЩИТНЫХ ОБОЛОЧЕК АЭС НА ЭТАПЕ СООРУЖЕНИЯ И ЭКСПЛУАТАЦИИ

ПРИМЕНЕНИЕ ЭКСПЕРТНОЙ СИСТЕМЫ ОЦЕНКИ НАПРЯЖЕННО-ДЕФОРМИРУЕМОГО СОСТОЯНИЯ ЗАЩИТНЫХ ОБОЛОЧЕК АЭС НА ЭТАПЕ СООРУЖЕНИЯ И ЭКСПЛУАТАЦИИ

Experimental study on the condensation of sonic steam in the underwater environment

Abstract Steam jet condensation is of great importance to pressure suppression containment and automatic depressurization system in nuclear power plant. In this paper, the condensation processes of sonic steam jet in a quiescent subcooled pool are recorded and analyzed, more precise understanding are got in direct contact condensation. Experiments are conducted at atmospheric pressure, and the steam is injected into the subcooled water pool through a vertical nozzle with the inner diameter of 10 mm, water temperature in the range of 25–60 °C and mass velocity in the range of 320–1080 kg/m2s. Richardson number is calculated based on the conservation of momentum for single water jet and its values are in the range of 0.16–2.67. There is no thermal stratification observed in the water pool. Four condensation regimes are observed, including condensation oscillation, contraction, expansion-contraction and double expansion-contraction shapes. A condensation regime map is present based on steam mass velocity and water temperature. The dimensionless steam plume length increase with the increase of steam mass velocity and water temperature, and its values are in the range of 1.4–9.0. Condensation heat transfer coefficient decreases with the increase of steam mass velocity and water temperature, and its values are in the range of 1.44–3.65 MW/m2°C. New more accurate semi-empirical correlations for prediction of the dimensionless steam plume length and condensation heat transfer coefficient are proposed respectively. The discrepancy of predicted plume length is within ± 10% for present experimental results and ± 25% for previous researchers. The discrepancy of predicted condensation heat transfer coefficient is with ± 12%.

Calcite formation on steamed concrete surfaces and its potential for sealing cracks

Under accidental conditions, concrete containments of nuclear power plants often remain as final barrier to the environment. To evaluate the contribution of self-healing processes of cracked concrete containments, small scale experiments were conducted for given accident scenarios to examine the mineralogical processes on cement crack surfaces in contact with hot steam-air mixes. The surface analyses with ESEM, µ-RFA and microscope showed a calcite formation, where layer thicknesses increased along the exposure times until reaching a stable level. The potential for crack healing was principally demonstrated through flow measurements, where the outflowing mass decreased during the experiments along with increasing calcite formation.

APPLICATION OF PASSIVE AUTOCATALYTIC RECOMBINER IN NUCLEAR POWER PLANT UNDER SEVERE ACCIDENTS

Passive Autocatalytic Recombiner (PAR) is a key safety equipment to mitigate the hydrogen risk in containment of Nuclear Power Plant (NPP). As we all know, PARs have been used to deplete hydrogen in Designed Basis Accidents due to the stable hydrogen depletion performance and flexible installation without power supply and manual operation. However, the application of PAR under Severe Accidents (SAs) have been got more and more attention. In some NPPs, PARs have been already required to be able to work in SAs. Therefore, the hydrogen depletion performance of PAR under SAs is very important for the design of hydrogen depletion system. In this work, first of all, the development of PAR research is overviewed briefly. Then the experiments, simulations or calculations under different conditions of SAs are described in detail, including the illustration of testing device or CFD model. The hydrogen depletion catalyst is also described from the perspective of catalyst preparation and characterization experiments. Finally, the future experiments or simulations which should be paid attention to are recommended.

The role of CFD combustion modeling in hydrogen safety management – VII: Validation for hydrogen deflagration in large-scale hydrogen-air-steam experiment

Abstract A postulated severe accident in a nuclear power plant (NPP) can lead to generation of large quantities of hydrogen in the NPP containment. Potential hydrogen combustion and ensuing dynamic pressure loads can be detrimental to the integrity of the containment and its structure, systems and components. This paper provides validation results of a computational fluid dynamics (CFD) modeling approach based on two combustion models, that is the Turbulent Flame-speed Closure (TFC) model and the Extended TFC (ETFC) model. The latter was further elaborated in order to investigate the influence of the laminar flame speed on the flame evolution and propagation modeling. As a result, a newly developed approach in form of the augmented ETFC model was introduced and validated as well. The hydrogen slow deflagration results from the Upward Flame Propagation Experiment (UFPE) conducted in the large-scale HYKA-A2 experimental facility have been used for validation. The main results of the performed validation are presented and discussed, with the focus on axial and radial flame front propagation, maximum pressure, pressure increase rate and flame front development.

Math hydrogen catalytic recombiner: Engineering model for dynamic full-scale calculations

To protect a hermetic enclosure and the equipment and systems of a reactor installation housed in it from damage caused by the ignition (explosion) of hydrogen, most nuclear power plants with pressurized water reactors are provided with a hydrogen concentration monitoring system and an emergency hydrogen removal system. These systems prevent the formation of explosive mixtures in the accident localization zone by maintaining the concentration volume of hydrogen in the mixture below the safety limits, which ensures the preservation of the density and strength of the hermetic enclosure and the operability of other localizing safety systems. A key component of an emergency hydrogen removal system is a passive autocatalytic hydrogen recombiner that operates on the principle of the catalytic recombination of hydrogen and oxygen.There is an urgent need for a full-scale dynamic calculation of the development of emergency conditions in a nuclear power plant containment accompanied by a large release of hydrogen. In efforts to achieve this, we constructed, and justified, a simple engineering thermohydraulic model of hydrogen removal in the operation of a passive autocatalytic recombiner based on the available experimental data.This paper presents the application results of the model as a part of contour industry codes RELAP, TRACE and CORSAR, intended, among other things, for carrying out multifactor and full-scale calculations of the dynamics of emergency processes with the release of hydrogen into nuclear power plant premises. This model allows us to substantiate the dynamics of local concentrations of gas components of a mixture in a confined space; the temperature of the mixture, the catalyst and the walls of the box; and the pressure when hydrogen or steam is supplied to the box.We have analysed various rates of hydrogen supply to a closed box to numerically substantiate the time at which the concentration reaches the maximum level. Furthermore, we have calculated the performance for several entrance concentrations of hydrogen, and obtained a satisfactory agreement between the dynamics of the concentrations, the temperatures of the catalyst and gas, and the productivity of the passive autocatalytic hydrogen recombiner. These calculations are based on the results of comparisons between calculated and available experimental data.

Study on mechanical behavior of containment in nuclear power plant during prestressing construction

Abstract Taking a third generation nuclear power plant as the research background, a refined model of a prestressed concrete containment which consisted of 270 complex spatial prestressed tendons was accurately established, using FE software ABAQUS. The cooling method was adopted to simulate the tensioning process of tendons. The simulation method of the tensioning of tendons during the prestressing construction of containment was verified by comparing with the measured data on site of the practical project. The mechanical behavior of the containment during the prestressing construction according to the prestressing construction scheme on site was studied in detail. Then the influence of steel liner and reinforcement on mechanical behavior of the containment during the prestressing construction of prestressed tendons was discussed. Finally, four prestressing construction schemes were analyzed and a reasonable optimization scheme was obtained.

Chemical Head Loss Measurements on Multi-Constituent Debris Beds in Small-Scale Experiments

A series of experiments on vertical head loss modules or columns to measure conventional and chemical head loss was carried out to support the resolution of Generic Safety Issue 191 for the Vogtle nuclear power plant (NPP). The head loss (conventional and chemical) was measured on multi-constituent fibrous debris beds of different particulate-to-fiber ratios (η). The debris beds were generated on a horizontal screen following the new procedure developed at the University of New Mexico and are summarized herein. The generated debris beds have been shown to produce repeatable and stable conventional head loss (CHL) and have the ability to detect chemical surrogates. Prototypical Vogtle NPP containment debris materials were used to form three different particulate-to-fiber–ratio (η) debris beds: 6.89 (thin bed), 2 (intermediate bed), and 1.15 (thick bed). The particulates were presented as 90% epoxy paint, 5% inorganic zinc, and 5% latent debris dirt by mass. The obtained results show that the measured CHL increased as the particulate mass increased in the debris beds. The average measured CHL values were 9.37, 6.4, and 5.66 H2O'' for η = 1.15, 2, and 6.89 debris beds, respectively. The debris beds with η = 2 and 1.15 were selected for the chemical head loss experiments.Standard aluminum (Al) chemical precipitates with specific batches were introduced to the head loss columns, and chemical head loss was measured. Precipitates prepared following the WCAP-16530-NP-A procedure [Lane et al., WCAP-16530-NP-A, “Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191,” Westinghouse Electric Company (2008)] or formed in situ by injecting metal salts under two different rates (0.75 and 7.5 mL/min) were tested. The results show that the thin debris bed (~10 mm) was more sensitive to the chemical precipitates prepared following the WCAP procedure compared to the intermediate debris bed (~25 mm) and thick debris bed (~55 mm). The measured chemical head loss was 0.35, 0.1, and 0.02 H2O''/mg of Al filtered by the debris beds. The in situ injection method has shown higher measured chemical head loss per unit mass of filtered precipitates than the WCAP surrogates for the debris beds of η = 2 (intermediate bed) and 1.15 (thick bed). Also, the results show a nonconclusive effect on the injection rate of metal salt to form in situ chemical precipitates on the measured chemical head loss.

Atucha-1 NPP containment venting analysis following SBO and LBLOCA events by GOTHIC code

Abstract Containment behaviour plays a key role in the safety framework of a Nuclear Power Plant (NPP). The GOTHIC thermal hydraulic code has been adopted to evaluate the Atucha-1 NPP containment responses during two postulated severe accident scenarios, Station Black Out and Large Break Loss of Coolant Accident without Safety Injection Pumps (SIPs), while assuming that the external cooling of the Reactor Pressure Vessel is carried out during the transients. The Atucha-1 NPP has a containment designed to work at full pressure, constituted by a steel sphere enveloped by a concrete shell, and having an annular gap of air in between. The target of the analysis is the evaluation of the effects caused by the additional production of steam in the reactor cavity as a consequence of the external vessel cooling, which could cause an increase in containment pressure, and lead to pressure values above the safety limit. The containment pressure and temperature, the distribution of hydrogen in the containment atmosphere and the water hold-up in the most relevant rooms have been analysed as target variables. Each accident scenario was simulated using two different nodalizations, characterized by a different level of refinement. The “detailed” nodalization is meant to be the most refined nodalization according to the available computational resources; having high fidelity three dimensional details, with a high number of cells. Taking into consideration that several sensitivities were performed, the “coarse” nodalization was developed in order to lower the demand for computational resources without significantly compromising the global scenario response. Both nodalizations are characterized by high complexity in the representation of rooms and their connections. Both accident transients, for each type of nodalization, were simulated for 200,000 s. At the end of the simulated transient, results showed that for the Large Break Loss of Coolant Accident pressure is predicted to surpass 5 bar, while the Station Black Out scenario is calculated to reach 4.4 bar. The performed sensitivities were simulated for 100,000 s and were meant to understand and characterize the impact of the different nodalization parameters (geometrical aspects, material properties, BCs). In addition, due to several code anomalies identified, several other sensitivity calculations were performed in order to find a way to analyse and mitigate the issues.

Hydrogen–Air–Steam Deflagration Experiment Simulated Using Different Turbulent Flame-Speed Closure Models

The paper presents the computational fluid dynamics (CFD) combustion modeling approach based on two combustion models. This modeling approach was applied to a hydrogen deflagration experiment conducted in a large-scale confined experimental vessel. The used combustion models were Zimont's turbulent flame-speed closure (TFC) model and Lipatnikov's flame-speed closure (FSC) model. The conducted simulations are aimed to aid identifying and evaluating the potential hydrogen risks in nuclear power plant (NPP) containment. The simulation results show good agreement with experiment for axial flame propagation using the Lipatnikov combustion model. However, substantial overprediction in radial flame propagation is observed using both combustion models, which consequently results also in overprediction of the pressure increase rate and overall combustion energy output. As assumed for a large-scale experiment without any turbulence inducing structures, the combustion took place in low-turbulence regimes, where the Lipatnikov combustion model, due to its inclusion of quasi-laminar source term, has advantage over the Zimont model.

Penetration depth of RC panels subjected to the impact of aircraft engine missiles

The aircraft collision with RC structures would lead to the global response and local damage, which are mainly caused by the impact of the entire aircraft and the massive engines, respectively. This paper mainly aims to study the penetration depth of RC structures, particularly for the nuclear power plant (NPP) containment, subjected to the impact of aircraft engine missiles. Firstly, the widely-used empirical formulae for predicting the penetration depth of rigid projectiles, as well as the existing notable engine missile impact test are introduced briefly. Then, by utilizing an advanced large-caliber air gas gun, a series of 1/10 reduce-scaled engine impact test on the 150 mm thick RC panels was conducted. The striking velocities were ranged from 94 m/s to 184 m/s, considering the taking off/landing and maximal cruising speeds of common commercial aircrafts, e.g., Airbus A320. The impact processes, missile deformation and panel damage were recorded and assessed. Finally, based on the present and existing tests data, the Haldar and Hamieh (1984) formula for rigid projectile was modified by introducing a reduction coefficient of 0.5 to predict the penetration depth of RC panels hit by the deformable engine, which is more accurate and practical than NEI (Nuclear Energy Institute) recommendation.

Method for evaluation of concrete containment structure subjected to earthquake excitation and internal pressure increase

SummaryThis study evaluates capabilities of the VecTor4 computer program to assess the seismic performance of a concrete containment structure subjected to both design‐basis and beyond design‐basis earthquakes. Based on a detailed nonlinear finite element model of the nuclear power plant containment structure, the seismic performance in terms of stress and strain distributions, cracks, yielding of reinforcement bars and tendons, as well as overall failure mechanism, is thoroughly reviewed. In addition, the post‐seismic performance of the containment structure subjected to a subsequent pressure increase is also investigated. Consideration is also given to the time‐dependent parameters of creep, shrinkage, and relaxation of prestressing tendons. It is found that the time‐dependent parameters and earthquake have a non‐negligible impact on pressure‐induced structural performance but that this impact can be predicted for a given earthquake. The proposed method is useful for assessing the progressive failure behaviour of containment structure under chain events.

Simulation of wall film condensation with non-condensable gases using wall function approach in component thermal hydraulic analysis code CUPID

In the containment of nuclear power plants, wall film condensation occurs with non-condensable gases under accident conditions. With non-condensable gases, condensation heat transfer on the containment wall can be degraded significantly because of the accumulation of non-condensable gases near the condenser wall; therefore, an investigation into the wall condensation heat transfer is of great importance to nuclear reactor safety. In this study, wall film condensation with non-condensable gases was simulated using the CUPID code. To evaluate the heat and mass transfer coefficients, a wall function approach was adopted to save the computational cost. To validate the model, a COPAIN condensation experiment was simulated using CUPID. The calculation results were compared with the COPAIN experiment data and results from the commercial CFD code (STAR-CCM+) results, which used the resolved boundary layer approach. From the comparison, good agreements were obtained between the CUPID code and the other results.

Post-benchmark simulation of the THAI+ facility TH27 experiment

ABSTRACT In order to maintain the integrity of a nuclear power plant containment and effectively manage a severe accident, it is necessary to understand phenomena occurring in the atmosphere of the nuclear power plant containment during the accident. A number of containment atmosphere mixing experiments have been performed in the dedicated experimental facilities, followed by the numerical simulations using lumped-parameter and computational fluid dynamics codes. This paper presents the THAI+ test facility experiment TH27 post-benchmark simulations performed with the lumped-parameter code ASTEC. The experiment TH27 was an initial operation test of the THAI+ facility, which has been recently constructed by expanding the experimental facility THAI with the newly constructed parallel attachable drum vessel. The experiment featured steam and helium injections and transport and mixing of gasses and steam between the two vessels, as well as wall heating and cooling of different vessels. The TH27 experiment was performed together with an international multistage benchmark, consisting of double-blind, blind, and open phases. The developed nodalization scheme and the features of the calculation are presented in the paper. The results of the calculations are compared to the experimental values for the main containment parameters – pressure, gas and wall temperatures, helium concentrations.

Large eddy simulations of the all-speed turbulent jet flow using 3-D CFD code GASFLOW-MPI

Turbulent flow at a wide range of Mach numbers may occur during design based accidents (DBA) or severe accidents in nuclear power plants (NPPs), for instinct, incompressible gas mixing flow (subsonic), gas mixture critical flow (transonic) or detonation (supersonic). All of them need to be simulated in safety analysis of NPP containments, therefore we need the computer code to have the capability to simulate all-speed flow. GASFLOW-MPI is a well-developed parallel all-speed CFD software used to predict gas turbulent mixing, chemical combustion and other related thermal-hydraulic phenomena in containments of NPPs. In order to validate the all-speed turbulent flow capability of GASFLOW-MPI, four turbulent jet cases over a broad range of Mach number have been performed in this paper. The large eddy simulation (LES) turbulent model is employed to capture more details of turbulence and flow features in this study. The standard Smagorinsky sub-grid scale model is utilized to model the unresolved turbulent behaviors at sub-grid scale. Both the instantaneous and time-averaged flow fields are analyzed as well as turbulent statistics to compare with the experimental data sets. The simulation results are in good agreement with the experimental data sets available in the literature. The shear layer instability is observed in the instantaneous flow field which leads to the end of the potential core and then the centerline velocity begins to decay. The size of the potential core and the decay rate of the centerline velocity are consistent well with the experiments. The turbulence intensity in the fully developed region agree well with the experimental data and the velocity power spectrum reveals the -5/3 law. Simulation results show that the compressibility has a significant effect on a free shear layer growing rate which is consistent with the experiments. The complex shock wave structure is also captured exactly by the second-order total variation diminishing (TVD) scheme. GASFLOW-MPI code has been successfully validated using the experimental data of turbulent jet flow at the wide regime of Mach numbers.

Tensile and shear bond strength between cement paste and aggregate subjected to high temperature

In the framework of studying the leakage rate of concrete containment of nuclear power plants during an accident, the mechanical characterization of interface between cement paste and limestone aggregate was performed with mechanical tests at the local scale, inside the laboratory of micromechanics and integrity of structures (MISt). The evolution of tensile and shear bond strength between cement paste and aggregate after heating was determined. It was revealed that the tensile bond strength was strongly affected by the temperature rise, with a decrease of about 71% at 60 °C. Moreover, the shear bond strength increased almost linearly with the increase of normal stress. According to the Coulomb criterion, the cohesion and the internal friction angle were determined. These parameters decreased after heating, which indicates occurrence of thermic damage between cement paste and limestone aggregate. Finally, it appears that the tensile bond strength is more affected by the thermal damage than the shear bond strength.

A CFD based modelling approach for predicting steam condensation in the presence of non-condensable gases

A typical post Loss-of-Coolant-Accident (LOCA) situation in nuclear power plant containments is a major safety concern. The simultaneous presence of steam, air and potentially hazardous hydrogen, which may get generated due to the metal-water vapour reactions, can lead to a singularly dangerous situation that may jeopardize the integrity of the containment structure. After steam leakage, wall condensation inside the containment is the primary heat transfer mechanism to maintain favourable safe pressure and temperature. Condensation of steam leads to strong, local, natural convective fluid flow currents, which are coupled with buoyancy and diffusional flow of the involved species. In this paper, an attempt has been made to address these complex transport processes, involving multicomponent gas species and phase-change phenomena, by CFD approach using Ansys CFX. A Wall Condensation Model (WCM) is utilized to handle wall condensation wherein the condensate is computed as a sink term. The simulated results are benchmarked against available data from several small and large scale experiments (such as TOSQAN, CONAN and others). These validation cases are chosen so as to cover different types of geometries. Specific experiments are also performed for studying the diffusion of helium jets into air filled volumes – such situations are also validated by solving the coupled conservation equations for the multicomponent mixture. Unless the net non-condensable gas (NCG) concentration is very low (less than about 6%), it is shown that the WCM, in tandem with multicomponent mixture model (MCM), works quite well for simulating the considered containment thermal-hydraulic situations. This provides the necessary confidence to further develop and validate the CFD based approach to, in the end, be able to reliably simulate post-accident thermal-hydraulics of large containment structures.