The development of predictive capabilities for the phenomena governing fluid flow and heat transfer in nuclear reactors encompasses a wide spectrum of methods and levels of detail, from using simple heuristic formulae or correlations to small-scale extremely complex computational models and algorithms. The objective of this paper is to demonstrate that it is not uncommon that specific existing modeling methods and assumptions which are taken for granted are not necessarily always correct, and their applicability to particular problems should be carefully examined and assessed. Also, selected issues are discussed concerning the importance of identifying appropriate spatial and temporal scales for a meaningful and consistent formulation of mechanistic models. Examples are shown illustrating the assessment and robustness of selected modeling assumptions using both experimental data of reference and parametric studies. These examples are based on the research performed by the author and his collaborators. Additional information about each example can be found in the references to the original publications on the corresponding issues.

The complexity of phenomena occurring during severe accidents in nuclear reactors, combined with a limited amount of experimental evidence, do not allow for formulating detailed reliable models or making highly accurate predictions of accident progression. Thus, the modeling consistency and a proper understanding of the uncertainties associated with the results of any computer simulations, including those caused by the imperfection and inherent limitations of the available experimental data used in model validation, are critical for improving accident mitigation capabilities and for enhancing the safety of current and future generations of nuclear reactors. The objective of this paper is to give an overview of selected issues illustrating the importance of: (a) identifying the dominant phenomena governing the progression of core meltdown accidents, and (b) formulating models which are consistent with our understanding of the underlying physics and chemistry and of the increasing level of randomness as the accident progresses. The results used as examples, in particular those pertaining to the research performed by the authors and their collaborators, have been obtained over past several years and documented in a several reports (in particular, in the USNRC NUREG series), but have never been included in copy-righted publications. The sources of any other experimental data used in the discussion of specific coupled experimental/modeling issues (typically, also reports of various agencies or labs) are clearly identified in the text. Since some of the examples include recent unpublished results of computer simulations performed using updated versions of the models which have already been published in other journals, only brief information about such models is presently shown, assuming that the reader can find details in the corresponding references.

Understanding the origins and propagation of flow-induced instabilities is very important for the operation and safety of the current and next generations of boiling water reactors (BWR), as well as for future new reactor types, such as the supercritical-pressure reactors, cooled by either water or CO2. Consequently, substantial (or even extensive) efforts have been made during the past 40+ years to investigate the physical phenomena governing such instabilities, and to develop models and computational tools to simulate the dynamics of marginally-stable/unstable boiling systems. The purpose of this paper is to present an updated perspective on, and discuss new aspects of, selected studies performed by the author and his collaborators, associates and graduate students, aimed at the development of various methods for the analysis of flow-induced instabilities in both boiling and supercritical-pressure systems. The impact of model and method selection on the results of predictions will also be discussed for a large class of systems, including the neutronically-coupled oscillations. Since the current paper is a part of the NED issue commemorating 40 years since the first NURETH?series conference, the list of references includes eleven (11) related papers published before in various NURETH proceedings, as well as other related journal articles and conference papers.

The existing evidence clearly shows that the physical phenomena governing gas/liquid two-phase flows are quite complicated even in the case of smooth conduits and simple geometries. Needless to say, the development of experimental, analytical and computational methods for predicting such flows in complex geometries is an even more complicated and challenging task. A configuration of interest to a broad range of industrial applications, including nuclear reactors, deals with the flow of two-phase mixture along the channels formed between narrow arrays of multiple parallel cylindrical elements (tubes or rods). The alignment of such elements is normally accomplished by installing spacer grids placed at regular distances along the flow. The presence of spacers actually affect flow conditions, including the velocity field, pressure drop, heat transfer and, in the case of two-phase flows, phase distribution. The objective of this paper is to present a comparative analysis of the results of a combined experimental, theoretical, and computational study of phase distribution around and downstream from complex-geometry spacer grids with split-vane type mixing devices. The main emphasis has been given to the analysis of the effect of proper interpretation of the experimental data on the modeling consistency. The importance of the understanding of uncertainties and limitations associated with the results of multidimensional computer simulations performed using mechanistic modeling principles based on an average bubble size is also discussed.

The multi-field two-fluid modeling concept has been successfully used in the past to perform Computational Multiphase Fluid Dynamics (CMFD) simulations of forced-convection boiling. Since such simulations are normally performed at the RANS-level, they do not allow for resolving local phenomena at the bubble-scale level. Instead, closure laws making use of empirical correlations deduced from selected experimental data are adopted. As different data sets of reference frequently correspond to different, and not always overlapping, experimental conditions, their application introduces considerable uncertainties into the predictions of the overall CMFD models. It has been shown before (Podowski et al., 1997) that one way to improve the accuracy of computational models is to apply a multiscale modeling approach, in which single-bubble-nucleation-level models, validated against separate-effect experiments, are used to formulate mechanistic RANS-level closure laws. This paper presents a mathematically rigorous and physically consistent model of the bubble ebullition cycle, which takes into account the coupling of transient heat transfer between solid walls and the surrounding coolant during wall quenching period. The proposed modeling approach covers a large spectrum of the geometries pertaining to both experimental boiling test sections and nuclear reactor fuel assemblies. The new model has been parametrically tested and validated against experimental data. A good agreement between the model predictions and experimental data has been obtained. The overall objective of the current work is to establish a first-principle theoretical framework for the formulation of mechanistic ensemble-averaged closure laws compatible with existing CMFD solvers.

The objective of this paper is to discuss selected two-phase flow and heat transfer issues which would have been of interest to my three recently deceased friends and colleagues: Geoffrey Hewitt, Bal Raj Sehgal and George Yadigaroglu. The main emphasis is on the importance of understanding the underlying physics for proper interpretation of experimental data and for the formulation of consistent and meaningful models at various levels of complexity.

The progress in multiphase science and technology has been driven to a considerable extent by the problems arising from nuclear reactor thermal-hydraulics. However, the needs for obtaining solutions to a variety of practical questions have had both pros and cons. The former have been reflected in the role of needs as an obvious research stimulation factor, the latter have been due to fact that the urgency of producing results often limited systematic investigations of the underlying physics. Consequently, the maturity of this field has been frequently questioned in the past, in particular as compared to the single-phase fluid mechanics and heat transfer. The objective of this paper is to present an overview of selected issues which, in the author's personal opinion, illustrate the current state of knowledge in regard to various aspects of multiphase physics and its application to reactor thermal-hydraulics (TH). The discussed issues include: theoretical fundamentals of multiphase fluid mechanics and heat transfer, formulation and limitations of closure laws, transition from correlations to models, consistency vs. accuracy (or validation vs. tuning), and challenges associated with the multicomponent/multiphase modeling of core degradation phenomena. Examples shown to illustrate selected issues come from several studies performed by the author and his collaborators and associates. They cover a broad range of scales, from micro-scale phenomena at, or below, the individual bubble level, to the scale of reactor fuel assemblies. Given the complexity of the multiphase thermal-hydraulics, the author is aware that a lot of room is still available for different opinions on the current state of affairs in this field. So, the overall goal of the paper is to stimulate constructive criticism and future discussions among members of the nuclear TH community.

Due to the practical importance of boiling heat transfer, various attempts have been made to date to formulate algorithms for computer simulations of heat transfer between vapor films and/or bubbles departing from heated walls and the surrounding liquid. However, due to problem complexity, the majority of testing, verification and applications of computational models have been limited to relatively simple problems, typically 1D or 2D. The purpose of this paper is to present the results of a study on the development of a novel approach to the complete three-dimensional (3D) modeling of simultaneous evaporation and condensation occurring at subcooled boiling flow conditions in heated channels. The focus of the proposed method has been on a multiple-option model of interfacial heat transfer, which allows for an accurate evaluation of the interfacial heat flux between vapor bubbles and the surrounding subcooled or superheated liquid. The modeling of a variety of heating/cooling situations has been made possible by using appropriate newly-developed expressions for a local, position- and bubble-shape-dependent, Nusselt Number at the bubble/liquid interface. To assure the required accuracy, the new heat transfer model was extensively tested and verified in a stand-alone fashion before it has been numerically coupled with a Level-Set (LS) type solver and implemented in the NPHASE/CMFD computer code. The complete model has been tested again and experimentally validated. The stand-alone tests included model verification against the theoretical solution of the Stefan problem, and the near-analytical solution to the model of time-dependent condensation rate around a 2D bubble. As a starting point to a three-dimensional analysis, a DNS-type model has been developed to evaluate the local Nusselt Number for subcooled/superheated liquid flow around bubbles of different (fixed) shapes. The results have been validated against the available existing data. The formulae for the local Nusselt Number which have been deduced as a result of the follow-up analysis have then been implemented in the overall model of evaporation/condensation heat transfer around flowing bubbles of varying shapes. The results of NPHASE/CMFD-based 3D simulations have been validated against experimental data for a single bubble, and a good agreement has been obtained. Subsequently, simulations for multiple bubbles at subcooled boiling flow conditions have been conducted to illustrate the wide range of applicability of the proposed approach, and to demonstrate the ability of the current computational framework to model micro-scale mechanisms behind complicated nonequilibrium boiling processes. It is envisioned that future uses of the first-principle methodology discussed in this paper as virtual experiments should facilitate the development and verification of various multiphase closure laws for RANS-scale models for a variety of boiling heat transfer situations. (C) 2018 Elsevier Ltd. All rights reserved.

The objective of this paper is to discuss the modeling principles of phenomena governing core degradation/melting and in-vessel melt relocation during severe accidents in light water reactors. The proposed modeling approach has been applied in the development of a new accident simulation package, COMPASS (COre Meltdown Progression Accident Simulation Software). COMPASS can be used either as a stand-alone tool to simulate in-vessel meltdown progression up to and including RPV failure, or as a component of an integrated simulation package being developed in Korea for the APR1400 reactor. Interestingly, since the emphasis in the development of COMPASS modeling framework has been on capturing generic mechanistic aspects of accident progression in light water reactors, several parts of the overall model should be useful for future accident studies of other reactor designs, both PWRs and BWRs. The issues discussed in the paper include the overall structure of the model, the rationale behind the formulation of the governing equations and the associated simplifying assumptions, as well as the methodology used to verify both the physical and numerical consistencies of the overall solver. Furthermore, the results of COMPASS validation against two experimental data sets (CORA and PHEBUS) are shown, as well as of the predicted accident progression at TMI-2 reactor. (C) 2019 Korean Nuclear Society, Published by Elsevier Korea LLC.

As a part of the integrated severe accident code development project in Korea, KAERI (Korea Atomic Energy Research Institute) has been developing a stand-alone severe accident analysis code, COMPASS (COre Meltdown Progression Accident Simulation Software), which simulates the in-vessel severe accident phenomena including the core heat up, material melting and relocation, corium behavior in the lower plenum, and vessel failure. For the purpose of COMPASS code validation, a numerical simulation was conducted for the PHEBUS FPT3 experiment, which is an integrated severe accident experiment, initiated in 1988 by IRSN. In the COMPASS code, the core and surrounding structure in a test section consist of a two-dimensional node system, and the mass and energy equations are established for the main component in the core and surrounding structures. The temperature evolution of the core and surrounding structure at various locations of the test section are pursued and compared with the experimental data as well as the numerical results of the MELCOR code. The hydrogen generation rate and final mass distribution are compared among the experimental data and the numerical results of COMPASS and MELCOR. Generally, the COMPASS and MELCOR codes predicted the temperature evolution and hydrogen generation rate during the experiment well. In addition, COMPASS predicted a slightly smaller mass relocation compared with the experimental data, whereas MELCOR has a larger mass relocation. (C) 2017 Elsevier B.V. All rights reserved.

(henceforth referred to as ''the Agreement'') for which this technical progress report has been prepared covers the performance and analysis of field experiments at the first TERESA plant, located in the Upper Midwest and henceforth referred to as Plant 0, and at two additional coal-fired power plants (Plants 1 and 2) utilizing different coal types and with different plant configurations. During this reporting period, all fieldwork at Plant 0 was completed. Stack sampling was conducted in October to determine if there were significant differences between the in-stack PM concentrations and the diluted concentrations used for the animal exposures. Results indicated no significant differences and therefore confidence that the revised stack sampling methodology described in the previous semiannual report is appropriate for use in the Project. Animal exposures to three atmospheric scenarios were carried out. From October 4-7, we conducted exposures to oxidized emissions with the addition of secondary organic aerosol (SOA). Later in October, exposures to the most complex scenario (oxidized, neutralized emissions plus SOA) were repeated to ensure comparability with the results of the June/July exposures where a different stack sampling setup was employed. In November, exposures to oxidized emissions were performed. Stage I toxicological assessments were carried out in Sprague-Dawley rats. Biological endpoints included breathing pattern/pulmonary function; in vivo chemiluminescence (an indicator of oxidative stress); blood cytology; bronchoalveolar lavage (BAL) fluid analysis; and histopathology. No significant differences between exposed animals and sham animals (exposed to filtered air) were observed for any of the endpoints; histopathological results are pending and will be reported in the next semiannual report. The scenarios evaluated during this reporting period were slightly modified from those originally proposed. We substituted a new scenario, secondary aerosol + SOA, to investigate the effects of a strongly acidic aerosol with a biogenic component. Since we did not observe any biological response to this scenario, the neutralized secondary aerosol scenario (i.e., oxidized emissions + ammonia) was deemed unnecessary. Moreover, in light of the lack of response observed in the Stage I assessment, it was decided that a Stage II assessment (evaluation of cardiac function in a compromised rat model) was unlikely to provide useful information. However, this model will be employed at Plant 1 and/or 2. During this reporting period, significant progress was made in planning for fieldwork at Plant 1. Stack sampling was carried out at the plant in mid-December to determine the concentration of primary particles. It was found that PM{sub 2.5} mass concentrations were approximately three times higher than those observed at Plant 0. In mid-February, installation and setup for the mobile laboratories began. Animal exposures are scheduled to begin at this plant on March 21, 2005. During the next reporting period, we will initiate fieldwork at Plant 1. At either or both Plants 1 and 2, a detailed Stage II assessment will be performed, even if no significant findings are observed in Stage I. The next semiannual report is expected to include a detailed description of the fieldwork at Plant 1, including toxicological findings and interpretation.

Core degradation phenomena in PHEBUS FPT-3 experiment has been simulated by using COMPASS code, which has been under developing in Korea to simulate severe accident scenario. In the test section of PHEBUS FPT-3 experiment, a fuel assembly consisting of 20 fuel rods and 1 control rod with a height of 1m has been installed within a surrounding in-vessel structure having 2 layers of shroud and pressure tube. Steam has been injected from the bottom of test section with a flow rate of 0.5g/s in a temperature of 165°C. The bundle power has been increased with time step by step in order to simulate the core degradation phenomena. Since the PHEBUS FPT-3 experiment had been served as the International Standard Problem in 2011, the initial and boundary conditions for the simulation has been obtained from the ISP-46 specification. In the PHEBUS FPT-3 experiment, thermocouples have been installed at the various vertical locations on a cladding surface, fuel rods, control rod and in-vessel structures. Most of the measured temperatures have been compared with the numerical simulation results, except the failed thermocouples. The cladding temperature has shown a steep temperature increase by the Zircaloy-steam oxidation about 10,000sec. In a COMPASS code, in order to simulate steam oxidation process, the Cathcart model has been used for the low temperature region and Baker-Just model for the high temperature region. For the comparison purpose, the numerical simulation by using MELCOR 2.1 has also been performed and the results of COMPASS and MELCOR 2.1 have been compared with the experimental data. Generally, the simulation results of COMPASS and MELCOR 2.1 are shown to have a good agreement with the experimental results. The temperature evolution patterns of numerical simulations are well predicting the thermocouple data, while the shroud temperature of MELCOR 2.1 is shown to have a little higher temperature compared with the experimental data. The hydrogen mass generation rate and the total generated mass of hydrogen are also compared. The oxidation starting time are a little bit earlier for the case of MELCOR 2.1 simulation, while the total generated mass of hydrogen are shown to have a good agreement between the numerical simulation and the experimental data. The final mass distributions by the mass relocation are also compared. COMPASS results are showing a smaller mass relocation compared with the experimental data, while MELCOR 2.1 are showing a higher mass relocation due to the earlier debris bed formation. In this stage, the numerical simulation has been performed only for the degradation phase of PHEBUS FPT-3 experiment, the analysis for the aerosol phase showing a fission product behavior are remained as the future works.

The use of supercritical carbon dioxide (SC-CO2) as a working fluid in energy conversion systems has many benefits, including high efficiency, compact turbomachinery, and the abundance of CO2. A very important issue for design optimization and performance analysis of future SC Brayton cycles is concerned with the SC-CO2 flow inside high-speed compressors and turbines. The objective of this paper is to present a novel modeling approach to, and its use in numerical simulations of, SC-CO2 flow inside a high-speed compact compressor. The proposed approach capitalizes on using three different physical and mathematical formulations of one-dimensional (1D) models, i.e., compressible and incompressible flow models using actual properties of SC-CO2 and a compressible ideal gas model, as a reference to verify the predictive capabilities of a three-dimensional (3D) incompressible flow model. The incompressible model has been used to perform simulations for a complete detailed multidimensional model of an SC-CO2 high-speed compact compressor. The advantages of the new model include numerical stability, computational efficiency, and physical accuracy. In particular, it has been shown that the model's predictions are consistent with selected published technical data.

The objective of this paper is to discuss the results of a combined physical and computational analysis of the dynamics of large deformable bubbles in conduits of different geometries and orientations. The situations analyzed included a vertical circular pipe and a narrow rectangular channel of different inclination angles. Three-dimensional simulations were performed at a DNS-scale using the PHASTA code combined with the level set method for interface tracking. The results of computer simulations for Taylor bubble flow in a vertical pipe have been verified against a simplified theoretical model and validated against available general evidence deduced from various experimental studies. The predictions for bubbles flowing along inclined rectangular channels have been validated against experimental data. Several modeling and numerical issues have been investigated, including the effect of a liquid microfilm between the bubble and the wall above it, and the impact of the blending region arising from the level-set model formulation on the accuracy of results. (C) 2015 Published by Elsevier Ltd.

To support the development of next-generation nuclear power reactors, new mechanistic based computational models of reactor thermal-hydraulics need to be formulated, tested, and implemented. A new subcooled boiling model has been developed with the objective of reducing dependence on empirical correlations. The model is able to accurately account for vapor generation on a heated wall as well as for condensation-induced heat and mass transfer from vapor to subcooled liquid. This model represents a substantial improvement over previous models and is part of an ongoing effort [1]. The new model has been extensively validated against experiments covering a wide range of fluids, geometries, and flow conditions. The accuracy of subcooled boiling models is dependent on a consistent formulation of diabatic multiphase flow. A complete formulation for a dispersed field model is presented, along with a consistent set of interfacial heat, mass, and momentum transfer models. The drag and virtual mass forces have been formulated using well-accepted expressions. Novel formulations of the turbulent dispersion force and a relationship for a near-wall-varying lift coefficient are presented. The complete subcooled boiling model has been implemented in the state-of-the-art computational multiphase fluid dynamics solver, NPHASE-CMFD. The experiments used in model validation cover flows of water, heavy water, and refrigerant. The geometrical configurations of heated channels include circular pipes, annuli, and rectangular conduits [2, 3, 4, 5]. The experimental conditions used in the present study correspond to a wide range of flow rates, pressures and heating rates. The same model has been applied consistently to all experiments of reference. That is, the model has not been calibrated to any specific experimental data set, and any adjustments in the modeling parameters between the simulations of different experiments have been made based on a priori analysis. This demonstrates the truly predictive nature of the presented model of subcooled boiling. Comparisons are shown between the predictions and experimental data for the axial and radial distributions of both vapor phase volume fraction and liquid subcooling. The observed agreement between the model and the experimental data is quite good for all simulated experiments.

This paper gives prediction to the transient heat transfer at Departure of Nucleate Boiling (DNB) point for subcooled flow boiling. The prediction is carried out by solving the heat conduction equations in cylindrical coordinates with convective boundary condition, which changes with the change of the heat transfer mode on the heated surface. DNB is assumed to happen at the complete dryout of liquid sublayer trapped between the heated wall and an elongated vapor clot, during the passing time of the vapor clot. Important parameters including initial thickness of the liquid sublayer, vapor clot length, vapor clot velocity and void fraction etc., are calculated from the Liu-Nariai model. The initial heater surface temperature is derived from the Jens- Lottes correlation. The transient changes of liquid sublayer thickness, surface temperature at DNB are reported. No obvious temperature jumping is observed at DNB. To predict temperate excursion at Critical Heat Flux (CHF), more simulations to the transient boiling and film boiling processes are needed.

Spacer grids in nuclear reactor fuel assemblies may have a significant effect on coolant flow and heat transfer within the reactor core. There has been a great deal of work carried out by the nuclear power industry to optimize spacer designs to improve reactor performance during both normal and accident conditions. Many of these designs employ mixing devices in order to enhance the heat transfer within the core. The effect of fuel rod spacer grids can be even more significant at multiphase flow conditions. The challenges associated with the modeling of two-phase flow and heat transfer with phase change in the spacer region are augmented by the complexity of the geometry that gives rise to both numerical and theoretical issues. The overall objective of the present study has been to develop a consistent multidimensional model and the corresponding solution methodology for the analysis of two-phase flows around mixing vanes in reactor subchannels. The proposed model has been implemented in a state-of-the-art computational multiphase fluid dynamics code, NPHASE-CMFD. This code uses a pressure-based finite volume solution method which is applied to RANS-level ensemble-average multifield equations of multiphase/multi-component fluids. It has already been extensively applied and tested before to simulate both adiabatic and diabatic gas/liquid flows in complex geometries [1], [2], [3]. This main focus the work reported in this paper has been on the PWR spacer design with Split-Vane type mixing devices [4]. The results of NPHASE-CMFD-predicted evolution of the velocity field and void distribution between the upstream and downstream sections around the spacer are presented. The results of parametric testing on the effect of inlet conditions and of the vapor-generation rate along the heated fuel rods are also shown, as well as those of model verification and validation against experimental data.

A general objective of this project was to develop, verify and validate mechanistic multidimensional models of local flow and heat transfer in supercritical carbon dioxide (S-CO2) devices and systems, and to demonstrate the application of the new models to selected components of S-CO2 nuclear energy transport systems. Both steady-state and time-dependent operating conditions have been investigated. The overall workscope consisted of the following three major parts: Development, testing and validation of a mechanistic model of forced-convection heat transfer in heated channels cooled using S-CO2 at slightly supercritical pressures; Development, testing and verification/validation of a new model of the dynamics of closed- loop S-CO2 heat transport systems; and, Formulation, testing and verification of a mechanistic model for the analysis of flow and pressure distribution in S-CO2 compressors. The results of the work performed for the project have been documented in several publications.

Fluids at supercritical pressures are considered to be very promising as coolants for Generation IV reactor systems, including the supercritical water cooled reactor (SCWR), the CO2-cooled reactor (SCCO2R) and the Brayton cycle as a secondary system in sodium-cooled fast reactors.The operating conditions of such reactors allow for a significant thermodynamic efficiency increase of the respective power plants. On the other hand, several unresolved issues must still be addressed in order to develop a viable design of supercritical fluid nuclear systems. Examples include: in-core heat transfer in SCWRs, heat exchangers in the S-CO2 Brayton cycle, and flows in complex geometries of SCO2 compressors.It has been shown before that the understanding of the effect of fluid property variations on turbulence is a major factor in our ability to predict the combined fluid mechanics and heat convection in systems and components using supercritical fluids. The objective of this paper is to present the results of analysis on the effect of local multidimensional flow and heat transfer phenomena on the temperature distribution inside future SCWRs. Two proposed SCWR designs have been considered: a single-pass and a two-pass coolant flow configurations through the reactor core.It should be noted that a preliminary documentation of the results of the current work has been included in the materials of the NURETH-15 conference.

The ability to predict the shape of gas/liquid interfaces is important for various multiphase flow and heat transfer applications [1, 2]. An issue of particular interest to nuclear reactor thermal-hydraulics is concerned with accurate predictions of bubble transport in subcooled boiling. The underlying include: nucleation at the heated wall, bubble transport away from the wall and vapor condensation in contact with subcooled liquid inside the reactor coolant channels. The purpose of this paper is to present the results of analysis aimed at using firstprinciple modeling principles based on directly tracking vapor/liquid interfaces to predict bubble shape and velocity evolution in heated fluids at subcooled boiling conditions. The uniqueness of the proposed approach is that a model is shown for the first time, capable of capturing the combined phase-change and kinematic phenomena governing bubble motion across a liquid layer experiencing a sharp temperature gradient from superheated to subcooled conditions. Thus, the effect of simultaneous evaporation and condensation can be investigated on the velocity and shape of vapor bubbles. The proposed approach is based on a modified level-set method, which has been implemented in the NPHASE-CMFD computer code. The coupled numerical solver can be used to simulate the evolution of gas/liquid interfaces in two-phase flows for a variety of geometries and flow conditions, from individual bubbles to free surfaces (stratified flows) [1, 3]. The emphasis in the present work has been on formulating an accurate computational model capable of capturing the nucleation-driven bubble growth at the wall, combined with a simultaneously occurring condensation at the tip of the bubble which is in direct contact with subcooled liquid. Such situations are particularly important in pressurized nuclear reactors (PWR) where local near-wall boiling may occur at locations along the flow where the liquid subcooling is still quite high. The issues discussed in the paper include: a description of the novel aspects of the proposed level-set-concept-based method, an overview of the NPHASE code modeling framework and a description of the coupling method between these two elements of the overall model. A particular attention is given to the consistency and completeness of model formulation for the liquid/gas interfacial phenomena corresponding to the coupled evaporation and condensation conditions of heat transfer. Also, the impact of the numerical assumptions and solution method on the accuracy and consistency of predictions will be discussed. The accuracy is measured in terms of the calculated bubble shape and size, and the gas and liquid velocity fields. The results of model testing and validation, including comparisons against analytic solutions for simplified yet physically meaningful situations, are also shown. CASL-U-2013-0103-000 The 15 International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-15 NURETH15-587 Pisa, Italy, May 12-17, 2013