Lumped Parameter Code Research Articles

Transient behaviour and heat transfer characteristics in debris beds: Simulation and analysis of the LIVEJ2 experiment

Multiple uncertainties still exist about the state of the debris in Fukushima Daiichi Nuclear Power Station (1F). In the past, the attention of the nuclear safety community was focused on the heat transfer characteristics in the case of an homogeneous pool, but little attention was given to address the melting and heat transfer in the presence of a debris bed constituted of materials with different melting points. This condition represents a challenge for CFD analyses, because it includes multi-physics conditions, such as a low melting point fluid convecting into a debris bed surrounded by a crust on the vessel wall which has received little attention compared to classical CFD analyses. Even though a comprehensive analysis of a related experiment (i.e. LIVE-J2) has been performed recently by Madokoro et al. (2023) little attention on the results has been paid to the effect of debris bed porosity and the existence of a gap between the vessel wall and the crust. In the paper we have modified the porosity resistance based on the Ergun equation and proposed a simple model for the gap conductance in the lower part of the crust. The results show an improvement in the prediction of the thermal stratification and the vessel temperature in the lower locations. In addition, highlight that such phenomena constitute key parameters to keep into consideration in the simulation of prototypical cases both for CFD and lumped parameter codes (e.g. MELCOR, MAAP).

Validation and Application of a Code for Three-Dimensional Analysis of Hydrogen–Steam Behavior in a Nuclear Reactor Containment during Severe Accidents

In a pressurized water reactor (PWR) during a loss of coolant accident (LOCA) or a station blackout (SBO) accident, water and steam are released into the containment building. The water vapor mixes with the atmosphere, partially condensing into droplets or condensing on the containment walls. Although a significant amount of water vapor condenses, it coexists with hydrogen generated by the reactor core oxidation. As water vapor condenses, the volume fraction of hydrogen increases, raising the risk of explosion or flame acceleration. As such, water vapor’s behavior directly affects hydrogen distribution. To conservatively evaluate hydrogen safety in a PWR during a severe accident, lumped-parameter codes have been heavily used. As a best-estimate approach for hydrogen safety analysis in a PWR containment, a turbulence-resolved CFD code called contain3D has been developed. This paper presents the validation results of the code and simulation results of hydrogen behavior affected by water vapor condensation and hydrogen removal by passive autocatalytic recombiners (PARs) in the APR1400 containment. The results provide insight into the three-dimensional behaviors of the hydrogen in the containment.

Analyses of Jet Buoyant Flow in a Multicompartment Containment Using an Open-Source Solver

Hydrogen mixing and distribution in a nuclear power plant containment during a postulated severe accident is a phenomenon with high safety relevance for current and advanced light water reactors. The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) SESAR Thermal-Hydraulics (SETH) project was carried out to generate experimental data on basic containment phenomena with the required instrumentation to assess computational fluid dynamics (CFD) and lumped parameter codes. In this paper, we analyze the PANDA SETH Test 16 of the OECD/NEA SETH project using containmentFOAM, which is a tailored open-source CFD code package for containment analysis developed at Forschungszentrum Jülich GmbH based on OpenFOAM.® SETH Test 16 addressed steam-air mixture transport driven by jets and plumes in the multicompartment PANDA facility (two vessels and an interconnecting pipe) at the Paul Scherrer Institut in Switzerland. Steam as a wall plume was injected at 0.040 kg/s and 140°C in vessel 1, which was initially filled with dry air at 108°C. The steam-air mixture was vented out from the top of vessel 2. The system pressure was constant at 1.3 bar for the whole test duration (4000 s). To simulate SETH Test 16, the k-ω shear stress transport turbulence model was adopted with the generalized gradient diffusive hypothesis buoyancy turbulence model. The simulation represented the experimental phenomenology reasonably well. The trends for the calculated temperature distribution were similar to those of the experiments and only yielded a slight discrepancy of about 2.7%. In addition, the distribution of the steam molar fraction was well predicted above the injection tube in vessel 1. However, below the injection tube in vessel 1, some discrepancies between the simulations and the experimental results were observed. These may have been caused by the higher turbulent mass diffusivity in the physical model and the challenge of the mesh resolution in the middle region of vessel 1.

Common OECD/NEA FIRE and PRISME Cable Benchmark Exercise

This paper deals with the common OECD/NEA FIRE and PRISME Cable Benchmark Exercise conducted for a realistic cable fire scenario in an electrical system of a nuclear power plant. The major goal was to simulate a real cable fire event in order to assess the behaviour of fire models for a complex and real fire scenario with the current state of knowledge. Since a numerical Benchmark Exercise on a real fire event is quite challenging, a three-step methodology was adopted: (1) a calibration phase of the cable fire models, (2) a blind simulation of a cable fire test, and (3) the real cable fire event simulation. The selected fire event from the FIRE Database occurred in 2014 in a heater bay of a nuclear power plant. Thirteen institutions from nine NEA member countries participated in this Benchmark Exercise. Simulations were performed using computational fluid dynamics codes, a lumped parameter code, and zone codes. The simulations have provided a wide range of results which reflect the difficulty for fire simulation codes to correctly predict the development of the fire heat release rate over time of a cable tray fire, even under simple, open atmosphere conditions.

CFD and LP code benchmark evaluating the onset of par operation in case of extremely low oxygen concentration

In many reactor containments Passive Autocatalytic Recombiners (PAR) are installed to mitigate the hydrogen threat in case of a severe nuclear reactor accident. PARs oxidize hydrogen catalytically into water vapor. The heat of catalytic reaction drives flow through the PAR unit by natural convection, which continuously brings “fresh” hydrogen-air mixtures to the PAR unit, thereby developing a self-sustained process until either hydrogen or oxygen falls below a minimum concentration.The test series (HR-33 to HR-35) of the OECD/NEA THAI-2 program investigated the influence of temperature, pressure and steam on the onset of hydrogen recombination in case of low or extremely low oxygen concentrations. These conditions with oxygen concentrations below 1 vol–% can occur in accident scenarios (e.g. late accident phase with molten core concrete interaction). A better understanding of the PAR performance in oxygen-limited conditions is important for hydrogen management in severe accidents.The test data of HR-35 was used for a blind benchmark exercise within the OECD/NEA joint project. Lumped parameter codes (MELCOR and COCOSYS) have been used as well as 3D/CFD codes (GASFLOW, GOTHIC and CFX). A variety of PAR models were used, including models based on engineering correlations, models with a detailed nodalization of the PAR unit, and a model with detailed heat and mass transfer process (REKO-DIREKT). Consequently, the benchmark provided a good opportunity to assess the PAR models that are deployed in the state of the art safety analysis tools, especially under conditions representative for the late phase of a severe accident.

On the validation of ATHLET 3-D features for the simulation of multidimensional flows in horizontal geometries under single-phase subcooled conditions

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

OPR1000 TI-SGTR 사고 MELCOR 해석의 자연순환유동 입력인자 생산을 위한 CFD 해석

We performed a 3-dimensional CFD analysis of a natural convective flow in the steam generator (SG) inlet plenum of an optimized power reactor 1000 MWe (OPR1000) to generate input parameters of the MELCOR code, a 1-dimensional lumped parameter code, which will be used in analyzing the temperature induced steam generator tube rupture (TI-SGTR) accident. The input parameters used in the MELCOR calculation will account for a 3-dimensional effect in the natural convective flow in the SG inlet plenum. To produce a credible CFD analysis result, we first established a CFD analysis methodology, which is developed from validation analyses against the steam generator design data of the OPR1000 and the Westinghouse 1/7 scaled-down test data, for the thermal mixing flow calculation between the SG and the reactor pressure vessel in the OPR1000. Though this CFD analysis, we produced the MELCOR input parameters, the recirculation ratio 1.96, thermal mixing fraction 0.79, discharge coefficient 0.13, and hot tube fraction 0.40, with an error range of approximately 10% to 25%. The input parameters will be used in the MELCOR analysis to evaluate a radioactive material release from broken SG tubes during the TI-SGTR accident.

Development of a lumped-parameter code for efficient assessment of in-vessel melt retention strategy of LWRs

Motivated by efficient assessment of the in-vessel retention (IVR) strategy employed in some designs of light water reactors, a lumped-parameter code named as transIVR was developed in the present study, which has the features as (i) quick estimates of transient one- and two-layer melt pool heat transfer with reasonable accuracies; and (ii) two-dimensional representation of heat conduction in the reactor pressure vessel wall, facilitating the coupled thermo-mechanical analysis of the reactor pressure vessel under thermal loads. The transIVR code was first benchmarked against the UCSB FIBS case of two-layer configuration and then validated against the LIVE-7V experiment. Both cases showed good predictions in the IVR-interested parameters like the heat flux profile, residual wall thickness and energy split. It indicates the code's capability in predicting two-layer and transient melt pool heat transfers, respectively. Finally, the code was coupled with ANSYS Mechanical for the efficient assessment of vessel integrity. Good results were achieved in the validation against the vessel failure experiment FOREVER-EC2. Hence, the transIVR code is an efficient tool suitable to address related safety concerns of light water reactors, including IVR efficacy and vessel integrity (with its coupling to mechanical solvers).

Simulation of hydrogen deflagrations with the lumped parameter code COCOSYS

In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated by the oxidation of zirconium in the fuel rod cladding with steam. If the hydrogen leaks into the containment of a pressurized water reactor or the reactor building of a boiling water reactor and if the design concept of passive autocatalytic recombining is not adequate or fails due to adverse atmospheric conditions, it might form a combustible mixture with air. In case of ignition, pressure peaks can occur that are relevant for the integrity of the containment or the reactor building and safety systems. In the past 20 years 45 hydrogen deflagration tests (HD tests) have been performed in the THAI/THAI+ test facility with a free volume of 60–80 m3 which is operated by Becker Technologies GmbH in Germany. The lumped parameter code COCOSYS V3.0 (Containment Code System) which is part of the software package AC2 2019 developed by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) gGmbH offers the models FRONT and COMB for the simulation of hydrogen deflagrations in the containment or reactor building of a nuclear power plant (NPP). COCOSYS simulations of select THAI+ HD tests are shown, and model development potential is deduced.

MELCOR Analysis of a SPARC Experiment for Spray-PAR Interaction during a Hydrogen Release

A series of experiments were performed in the SPARC (spray-aerosol-recombiner-combustion) test facility to simulate a hydrogen mitigation system with the actuation of a PAR (passive auto-catalytic re-combiner) and spray system. In this study, the SPARC-SPRAY-PAR (SSP1) experiment is chosen to benchmark the MELCOR (a lumped-parameter code for severe accident analysis) predictions against test data. For this purpose, firstly we prepared the base input model of the SPARC test vessel, and tested it by a simple verification problem with well-defined boundary conditions. The implementation of a currently used PAR correlation in MELCOR is shown to be appropriate for the simulation of a PAR actuation experiment. In an SSP1 experiment, the PAR is reacting with hydrogen, and the spray actuation starts as soon as hydrogen injection is complete. The MELCOR simulation well predicts the pressure behavior and the gas flow affected by operating both a PAR and spray system. However, the local hydrogen concentration measurement near the inlet nozzle is much higher than the volume average-value by MELCOR, since high jet flow from the nozzle is dispersed in the corresponding cell volume. The experimental reproduction of the phenomena we expect, or, conversely, the identification of phenomena we do not understand, will continue to support the verification of analytical models using experimental data and to analyze the impact of spray on PAR operations in severe accident conditions.

Development and validation of a new module in the ATHROC for simulation of aerosol particles removal by spray in containment

As the main carrier of fission products, aerosols will transport in the containment and even leak into the environment during accident conditions, resulting in radioactive pollution. Thus, it is of great significance for radiation protection to study the aerosol distribution mitigation measures in containment after severe accidents. A multi-compartment lumped parameter code named ATHROC (Analysis of Thermal Hydraulic Response of Containment) was developed to analyze the thermal–hydraulic behaviors in the containment after accidents for China advance PWR. In order to satisfy the analysis of aerosol behavior in containment, an aerosol module based on the aerosol dynamic and hygroscopic models was added to the code and coupled with spray removal mechanism models. Three different types of benchmark experiments were selected for assessment of the improved code, including CVTR, LACE and CSE tests. Most of the predicted results of thermal–hydraulic responses and aerosol distribution are in realistic agreement with the experimental data. The comparative results show that the code has capabilities to accurately predict the thermal–hydraulic responses during spray activation, aerosol distribution and removal of aerosols by spray droplets. And the problems identified in the code will be further optimized in the future. Currently, the developed improved code satisfies the analysis of aerosol behavior in containment and provides a design for aerosol removal for spray system arrangement.

Modeling of wall condensation in the presence of noncondensable light gas

During a loss of coolant accident in nuclear reactor, significant amounts of steam and hydrogen can be released in the containment. Condensation of steam in the presence of noncondensable gases on the containment walls and structures is a key issue because of its role in removing heat from the atmosphere. Extensive experimental and theoretical studies have been carried out for a better understanding of this complex phenomenon which involves several physical processes and parameters. When condensation takes place in the presence of noncondensable gases a liquid film is formed and noncondensable gases accumulate at the interface. The diffusion of steam through the gaseous layer depends on the gas composition, velocity, temperature and pressure. The formation of a gaseous layer leads to a significant reduction of heat transfer. Simulations based on CFD or lumped parameter (LP) approaches use correlations to estimate heat and mass transfer due to condensation. In this work, we are interested in theoretical correlations based on the diffusion layer theory using heat and mass transfer analogy. Among the variables and parameters affecting wall condensation, the effect of the gas mixture properties in particular the diffusion coefficient modeling is investigated. Test cases of steam injection into an enclosure filled with air or air-hydrogen mixture are simulated with a LP code using two different correlations for the evaluation of heat and mass transfer. Each correlation is based on different formulations of the effective diffusion coefficient. The ISP47 test performed in the MISTRA facility was used for validation purpose. The results showed that the addressed heat and mass transfer correlations underestimate the condensation rate in the steady state. Furthermore, the impact of the effective diffusion coefficient modeling on the heat and mass transfer turned out to be significant compared to the effect of the heat and mass transfer correlation.

Simulation of LOCA-typical containment conditions with COCOSYS on the basis of THAI-test TH-29.3

Abstract In case of a postulated loss-of-coolant accident (LOCA) in a light water reactor, core degradation might occur if emergency measures fail. During this process a large amount of hydrogen might be generated by the oxidation of the fuel rod cladding. If the hydrogen leaks into the containment of a pressurized water reactor and if passive autocatalytic recombining fails, it might form a combustible mixture with air. In case of ignition, pressure peaks can occur that are relevant for containment integrity. During LOCA the third relevant component of the containment atmosphere is steam. Test TH-29.3 aims at expanding knowledge regarding the interaction of steam and light gas (e.g. hydrogen) and was conducted in the THAI test facility. During the test steam and helium (as ta substitute for hydrogen) are injected into the test vessel. A steam-helium cloud expands downwards and a helium-rich layer forms at the boundary of the cloud. The test is simulated by the use of the lumped parameter code COCOSYS (Containment Code System) which itself is part of the software package AC2 developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH.

Validation of the lumped parameter code COCOSYS against large‐scale OECD PRISME 2 fire experiments

SummaryFire safety analysis is a major issue for nuclear power plants (NPPs) in the context of deterministic safety assessments as well as of probabilistic safety analyses. Oil reservoirs and cables represent major fire loads. Therefore, simulations of oil and cable fires are of interest for quantifying the risk of such internal hazards in NPPs. To investigate the applicability of lumped parameter (LP) modelling, validations against fire experiments are required. In this way, results obtained with the LP code COCOSYS for simulations of oil and cable fire experiments conducted in the OECD PRISME 2 Project are presented. The PRISME 2 VSP (vertical smoke propagation) tests involving oil fires in a confined and mechanically ventilated facility were used to assess the ability of the LP code to simulate smoke propagation through a horizontal opening from the fire compartment to a compartment on top of it. As it was already identified in the “International Collaborative Fire Modelling Project (ICFMP),” this type of opening might cause problems in fire simulations, particularly for zone or LP fire models. In these simulations, attention has been paid to the coupling between the fire and the surrounding environment due to the decrease of oxygen concentration. Furthermore, different cable materials have been tested in the PRISME 2 CORE (completing and repeating) test campaign. By simulating the CFS‐3 (cable fire spreading) test with confined underventilated conditions, the applicability of the COCOSYS cable fire model with input parameters deduced from open atmosphere fire tests (CORE‐2) was analysed. Results show that the applicability of a LP fire model to predict the pyrolysis rate is partly limited for both oil and cable fires, in confined environment. However, simulations with prescribed pyrolysis rates show encouraging results in good agreement with the experimental data and underline the capability of the LP code COCOSYS to simulate the interaction between the thermal hydraulics inside compartments and the fire source.

A Review of the CFD Modeling Progress Triggered by ISP-47 on Containment Thermal Hydraulics

The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency International Standard Problem 47 (ISP-47) was aimed at assessing the predictive capabilities of computational fluid dynamics (CFD) and lumped-parameter codes regarding hydrogen mixing under representative thermal-hydraulic conditions of a loss-of-coolant-accident. The benchmark consisted of two systematic steps. The first step was a fundamental model assessment based on quasi-steady-state separate-effects tests in the French TOSQAN facility (7 m3, IRSN, Saclay) and MISTRA facility (100 m3, CEA, Saclay) regarding steam condensation, buoyant turbulent flows, and mixed atmospheric conditions. The second step was based on a more realistic experimental transient in the multicompartmented German Thermal-hydraulics, Hydrogen, Aerosols and Iodine (THAI) facility (60 m3, Becker Technologies, Eschborn). At that time, the blind and open analysis revealed that CFD codes needed further improvement regarding modeling of turbulence in buoyant flows, steam condensation, temperature and species concentration, and stratification buildup as well as their dissolution. This result triggered a comprehensive experimental and analytical effort, e.g., within the German national THAI, the OECD-THAI, and the OECD-SETH-1 and OECD-SETH-2 projects. Now, 10 years later, this paper aims to benchmark the state-of-the-art containment CFD model, developed at Forschungszentrum Juelich and RWTH Aachen University, and to highlight the progress made and the remaining open issues.

Benchmark exercise TH-27 on natural convection with steam injection and condensation inside the extended THAI facility

An international double-blind and blind code benchmark was conducted in the frame of the German THAI program. The basis for the benchmark was test TH-27 which was performed as the commissioning test for the extended, two-vessel THAI+ test facility. The test provides code validation data for simulating flow and mixing inside a containment test facility under typical accident conditions which are of practical interest to nuclear reactor safety. The TH-27 benchmark addresses the phenomena of gas mixing, steam condensation and stratification behavior of light gases under typical containment flow conditions.Double-blind, blind and open simulations were performed by thirteen participants using lumped parameter (LP) models (ASTEC, COCOSYS, MELCOR) as well as CFD codes, namely FLUENT, CFX, GASFLOW and GOTHIC. The challenge of the double-blind benchmark was to generate a model and simulate a long lasting transient without previous model calibration based on knowledge gained from earlier tests being performed in the facility. The experimental measurements allow quantifying transport mechanisms and flow conditions between the two vessels as well as inhomogeneities of the gas mixtures of the vessels. Overall, the TH-27 benchmark demonstrated the high prediction quality of LP and CFD codes provided that dedicated nodalization guidelines are considered. On the other hand the benchmark revealed noticeable user influence for LP codes, especially due to different nodalization schemes and partial lack of special treatment regarding plumes or upward directed jets.

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.

Three-dimensional CFD analysis of hydrogen-air-steam explosions in APR1400 containment

The present study concerns the three-dimensional CFD analysis of hydrogen explosions in APR1400 containment. Initial conditions for combustion simulations are obtained by a cost-effective coupling with a lumped parameter code for preceding mixture distribution analysis. Three severe accident scenarios are examined: Small-Break Loss-Of-Coolant Accident (SB-LOCA) and Station Black-Out (SBO) with activated as well as deactivated three-way valve. Only in the last case, thermal ignition of the hydrogen-air-steam mixture could be triggered. Flame propagation in the containment and particularly in the In-containment Refueling Water Storage Tank (IRWST) is investigated in detail. Additional generic scenarios demonstrate the methods ability to reproduce strong Flame Acceleration (FA) and even the hazardous Deflagration-to-Detonation Transition (DDT).

Parametric study on density stratification erosion caused by a horizontal steam jet interacting with a vertical plate obstruction

During postulated severe accident scenarios in nuclear power plants, a hydrogen-rich layer might form at the top of the reactor containment. Various flow patterns resulting from the release of steam from the primary circuit might break the layer and redistribute hydrogen in the containment. The prediction of the gas transport during the accident requires detailed modeling of the processes involved. Advanced lumped parameter codes or computational fluid dynamics codes are used for this purpose. These codes need to be validated against experimental data obtained in large scale experimental facilities, where scale distortions are reduced. In order to obtain the required data with high spatial and temporal resolution, experiments were carried out in the PANDA facility in Switzerland as a part of OECD/HYMERES (HYdrogen Mitigation Experiments for Reactor Safety) project. The present experiments address the breakup of a layer rich in helium (used as simulant for hydrogen), under steam environment and its redistribution in two interconnected vessels (total volume of 183.3m3) under the action of a diffused flow resulting from the interaction of a horizontal steam jet with a vertical plate obstruction. The influence of the distance between the jet exit and the obstruction on the flow pattern was investigated. Spatial and temporal distribution of the gas concentration, the temperature and local gas velocity field were measured. It was found that a small change in the geometric configuration lead to a large change in the flow pattern. Reducing the jet-obstruction distance slowed down the helium-layer erosion process by a factor of two. Additionally, the creation of a concentration stratification in the adjacent vessel connected by an interconnecting pipe was observed.

Evaluation of Passive Containment Cooling with an Advanced Water Film Model in a Lumped-Parameter Code

Falling water films have been employed for passive containment cooling in several Generation III pressurized water reactor designs. In this paper, the lumped-parameter (L-P) containment code system COCOSYS with an advanced water film model is applied to evaluate the performance of a passive containment cooling system (PCCS) during accidents. Based on the recent work and with further modification, an integrated water film model is developed. The new model considers different flow regimes of a liquid film as it flows downward and is being evaporated. The integrated model has been adapted to the L-P code and then implemented into COCOSYS. The new model enables the containment code to capture previously neglected phenomena, including the behavior of film breakup due to the reduction in mass; the formation of rivulets; the change in coverage rate and the development of rivulets; the change of velocity distribution as well as film thickness by considering the interfacial shear stress created by countercurrent air on the film surface; the hysteresis of rivulets, i.e., the process of advancing or retreating, involving changes in contact angles; and the influence of waves on the film surface.The new model is validated against existing test results and experimental observations in the authors' recent work and is further modified in this paper taking into account the influence of waves and the processes of rivulet hysteresis. The model is then assessed based on test nodalization, and the expected phenomena are observed. Afterward, the new model is applied to evaluate the performance of PCCS film cooling employed in the AP1000 containment.It is concluded that the original film model tends to underestimate the pressure loads due to the absence of film breakup, rivulet behavior, and shear stress models. The coverage rate, as a new factor captured in the new model, limits the evaporation rate and thus restricts the cooling efficiency of the falling film. The sensitivity analysis reveals that the contact angle and hysteresis phenomenon, which were not previously considered in the code, play significant roles in PCCS film cooling. The advancing contact angle of the rivulets is a decisive factor for the peak pressure, while the retreating contact angle is influential in the later phase of cooling. It can be inferred from the study that the ideal situation for PCCS cooling is that in which the water film is approaching complete dryout at the bottom of the containment. The newly developed liquid film model helps improve the accuracy and reliability of the simulation results.