Liquid Metal Coolant Research Articles

Pulsating Heat Pipe Performance Modeling with Liquid Metal Coolants Under Hypersonic Aerothermal Heating

Hypersonic heating loads concentrate at leading-edge compression areas to create excessively high local temperatures and thermally driven stresses. The fast and reliable thermal dispersion of heat pipes with significantly high thermal conductance can alleviate these localized thermal stiffness problems. The pulsating heat pipe (PHP) holds numerous advantages when compared to capillary, constant-conductance heat pipes for hypersonic thermal management applications, primarily because they lack a wicking structure. This paper numerically investigates thermal performances of a four-turn C103 niobium alloy PHP operating with lithium, potassium, sodium, and a eutectic sodium–potassium alloy (NaK-78) when exposed to heating conditions relevant to the hypersonic environment, investigating flight Mach numbers ranging from 6 to 8 and dynamic pressures ranging from 40 to 44 kPa. The robust thermofluidic properties of liquid metals, along with the powerful fluid pulsation induced in the PHP, can provide significant thermal transport from hot stagnant regions of the leading edge to the cooler trailing surfaces. Potassium showed superior thermal performance when compared to other liquid metal coolants under the presently tested conditions, with overall PHP thermal conductance as high as 34 W/K predicted for Mach 8 flight conditions. In contrast, sodium was associated with startup difficulties in the PHP; this paper attributes this to its significantly larger thermal conductivity, which can limit the vapor pressure difference over the liquid slug lengths. These predictions indicate an overall latent heat transfer dominance of around 70–95% in liquid metal PHPs.

A self-organized sandwich structure of chromium nitride for ultra-long lifetime in liquid sodium.

The development of fast neutron reactors with improved efficiency and sustainability, being a tangible solution to the large-scale utilization of nuclear energy, serves as a critical step prior to the commercialization of fusion energy. These reactors use liquid metal coolants, which can weaken the durability of metallic components. Conventional design of protective coatings counts upon thermodynamics, which often overlooks the kinetic factors such as structural evolutions, resulting in deteriorated coating properties. Herein, we present a novel interface-engineering strategy involving the control of the phase transformation direction and interface diffusion reaction. Through iterations of self-organization, desired surfaces and interfaces can be achieved for materials used in harsh environments. Specifically, a CrN-coated steel sample with an interfacial Cr layer was designed and fabricated. After ultra-long (up to 6000 h) immersion in liquid sodium, the CrN/Cr coating structure was converted into a sandwich Cr2N/CrN/Cr2N structure dynamically. As a consequence, the coating system exhibited enhanced properties, namely increased surface hardness (by ∼36%), reduced coefficient of friction (by ∼13%), and enhanced interfacial adhesion (by ∼37%). Thus, the proposed strategy can guide the future design of robust coatings with ultra-long service life in harsh environments.

MTIR-SCP nuclear safety concept

The MTIR-SCP reactor is a research reactor with a light-water coolant having a fast neutron spectrum in nominal operation. None of the existing or projected reactors has a set of these features, therefore, special attention should be paid to the choice of the design of control rods and their location. The article considers various designs of control rods characteristic of the VVER-SCP power reactor currently under development, as well as existing research reactors with a liquid metal coolant having a fast neutron spectrum, taking into account the features of the MTIR-SCP: a seven-element control rod assembly design, as well as a ring-type absorbing element. Efficiency calculations are carried out for each of the structures under consideration and it was determined that the design of a control rod with an annular type of absorbing element has the highest specific efficiency. This design is chosen as the main one for the MTIR- SCP reactor. Criteria are proposed and optimization of the placement of control rods in the reactor core is carried out in order to reduce the uncertainty of obtaining neutron-physical characteristics in the central autonomous loop channel and increase the flow density in it. When optimizing the placement of control rods in the research reactor, their influence on the main neutron-physical characteristics, and especially on their distribution during operation under nominal conditions, is taken into account. Emergency protection and compensating rods that compensate for temperature and density effects when the reactor reaches the nominal power level have been removed, therefore their effect on the neutron-physical characteristics of the reactor and the loop channel during operation at power is not significant, and the compensating rods are located in the core and are removed as the fuel burns out, therefore their effect on the spatial The distribution of neutron-physical characteristics will be significant. To reduce this effect, burnout compensators and automatic regulators are proposed to be moved outside the core into the reflector. For various reactor conditions, the efficiency of the control and protection system was evaluated. It is shown that in various states, the control rods transfer the reactor to the required level of subcriticality, in accordance with NP-009-17.

CorTAF-LBE: A full scale subchannel-level thermal-hydraulic characteristics analysis code for LBR core based on OpenFOAM

The lead-bismuth cooled fast reactor core adopts box-type design, and the fuel rod bundle is fixed by wrapping wires. The complex geometric structure of the fuel assemblies significantly affects the flow and heat transfer status of the liquid metal coolant. Building upon the open-source CFD platform, OpenFOAM, this work established conservation equations for liquid metal coolant, fuel rod thermal characteristic models, inter-wrapper coolant analysis models, and coupled heat transfer models, and developed the subchannel level thermal-hydraulic analysis code CorTAF-LBE for whole-core analysis of LBR core using the finite volume method. The wire-wrapped rod bundles internal flow and heat transfer experiments of lead-bismuth, carried out based on the THEADES and KYLIN-II circulation loops, were selected to perform verification and validation. The results match well with the experimental data and CFD fine-modeling prediction. The steady-state numerical simulation of three wire-wrapped rod bundle assemblies with inter-wrapper flow was carried out under normal operation condition. The temperature and pressure distributions of the coolant in each subchannel and inter-wrapper flow region, the temperature of cladding and other key thermal-hydraulic parameters were obtained. The role of the inter-wrapper coolant in heat removal from the fuel assemblies was revealed. The work in this paper has referential value for the optimization of lead-bismuth cooled fast reactor design and safety assessment during operation.

Changes in chemical composition of TixAl1−xN coatings immersed in oxygen-saturated Lead–Bismuth Eutectic at low and moderate temperatures (250 °C ≤ T ≤ 410 °C)

Lead–Bismuth Eutectic (LBE) will serve as a liquid metal coolant in accelerator-driven systems, posing a corrosive threat to exposed materials. To address this challenge, TixAl1-xN coatings (0.38 ≤ x ≤ 0.58) were applied onto AISI 316 L austenitic stainless steel using the reactive bipolar magnetron sputtering technique. These coated samples underwent immersion in static oxygen-saturated LBE at temperatures of 250 °C and 360 °C for durations of 500 h and 1000 h, and at 410 °C for 500 h. XPS depth profiles revealed minimal oxidation at 250 °C even after 1000 h. However, at 360 °C, a mixed oxide layer of (Ti, Al)Ox was formed on the coating's surface. Exposure to LBE at 410 °C for 500 h led to the creation of an oxide bilayer comprising TiO2 (outer sublayer) and (Ti, Al)Ox depleted of Ti (inner sublayer). A model is proposed to illustrate the oxidation mechanism.

Thermodynamic Analysis and Comparison of Power Cycles for Small Modular Reactors

Small nuclear power plants can provide a stable, carbon-free energy supply to civil infrastructure and industrial enterprises in remote regions isolated from unified energy systems. More than 70 projects of small modular reactors are currently being developed by IAEA member countries; several low-power power units are already supplying thermal and electrical energy to consumers. One of the main limitations standing in the way of widespread dissemination of this technology is the high specific capital cost of a low-power nuclear power plant; therefore, new scientific and technical solutions are needed in this industry. Increasing the thermodynamic efficiency of power cycles of small modular reactors can become a driver for reducing the cost of supplied electrical energy. This paper presents the results of a comprehensive thermodynamic analysis of existing and promising power cycles for small modular reactors. In addition to traditional steam power cycles, cycles using non-traditional working fluids, including carbon dioxide, freons, and helium cycles, are considered. Optimal sets of thermodynamic parameters were determined to ensure maximum net efficiency of electricity production. For water-cooled reactor plants, a maximum efficiency of 33.5% at an initial temperature of 300 °C could be achieved using a steam turbine cycle. It was revealed that for reactor plants with liquid metal and liquid salt coolant in the range of initial temperatures above 550–700 °C, the maximum thermal efficiency was provided by the Brayton recompression cycle with a carbon dioxide coolant: the net electrical efficiency exceeded the level of steam turbine plants, with intermediate superheating of the steam, and could reach a value of 49.4% at 600 °C. This makes the use of these cycles promising for low-power nuclear power plants with a high initial temperature. In small gas-cooled reactor plants with a helium coolant, the use of a binary cycle consisting of a helium Brayton cycle and a steam-powered Rankine cycle provided an efficiency of 44.3% at an initial helium temperature of 700 °C and 52.9% at 1000 °C. This was higher than in the Brayton cycle with a recuperator, with a minimum temperature difference in the heat exchanger of 20 °C: the efficiency was 40.2% and 52%, respectively. Also, the transition to power cycles with non-traditional working fluids will lead to a change in the operating conditions of turbomachines and heat exchangers.

Reaction Capsule Design for Interaction of Heavy Liquid Metal Coolant, Fuel Cladding, and Simulated JOG Phase at Accident Conditions

High temperature corrosion of fuel cladding material (15-15Ti) in high burn-up situations has been an important topic for molten metal-cooled Gen-IV reactors. The present study aims to investigate the simultaneous impact of liquid lead (coolant side) and cesium molybdate (fuel side) on the cladding tube material. A capsule was designed and built for experiments between 600 °C and 1000 °C. In order to simulate a cladding breach scenario, a notch design on the cladding tube was investigated pre- and postexposure. Material thinning by corrosion and leaching at temperatures ≥ 900 °C caused breaches at the notches after 168 h exposure. The temperature dependent cladding thinning phenomenon was used for kinetic interpretation. As the first of a two-part study, this paper will focus on the exposure capsule performance, including metallographic cross-section preparation and preliminary results on the interface chemistry.

Atomistic modelling of the Pb-Bi system from a constructed bond order potential

Lead-bismuth eutectic (LBE) alloys are currently considered one of the most promising liquid metal coolants for fast reactors, and especially accelerator-driven systems (ADS). An analytical bond-order potential (BOP) of Pb-Bi has been constructed to account for the electronic interactions and angular contribution of Pb-Bi system. The newly developed BOP has been validated to show a better performance than the Pb-Bi potentials already published in the literature. Molecular dynamics simulations based on current BOP has been adopted to investigate various properties of Pb-Bi binary system. It accurately predict the heats of formation, density, and thermal expansion coefficient of liquid LBE at 700 K. The diffusion coefficient and viscosity of liquid Pb44.5Bi55.5 are calculated, which are in good agreement with experimental observations in the literature. These findings provide a foundation for future design of Pb-Bi eutectic alloys for use in ADS and heavy-metal-cooled fast nuclear reactors.

Effects of different heat transfer fluids on thermal distribution and electrochemical performance of PEMFC with a non-isothermal multiphase model

Proton exchange membrane fuel cell (PEMFC) with long service life, low parasitic loss and high-power density is one of the main technologies to achieve zero emission of automobiles. Efficient coolant is crucial for reducing the volume and complexity of thermal management system. In this study, the 3-D non-isothermal multiphase PEMFC model coupled with parallel cooling channels was established by using CFD simulation platform, and then the effects of four heat transfer fluids, namely deionized water, Syltherm800 heat transfer oil, 1.0 vol% Al2O3 nanofluid and K0.78Na0.22 liquid metal, on the heat and mass transfer characteristics and electrochemical performance of PEMFC were investigated. The results show that with the Re number increases, the PEMFC maximum temperature, temperature difference and index of uniform temperature (IUT) of water, nanofluid and liquid metal cooling modes decreased significantly than those of heat transfer oil cooling mode. While the heat transfer oil cooling mode has the lowest IUT and the highest peak temperature because of more oxygen involved in the side reaction. The relatively low temperature in PEMFC with liquid metal cooling mode will increase the water content of the membrane and improve the proton conductivity. In addition, at Tin = 345 K and Re = 900 condition, the net power of PEMFC with liquid metal cooling mode is up to 128.23 W, which is 1.62 % higher than that of water cooling mode. This confirms that the application of liquid metal coolant in thermal management system can effectively improve the performance of PEMFC.

Adaptation of the turbulence model to the heat transfer in a rod bundle

Fast neutron reactors using liquid metal or gas coolants are one of the most actively studied fields of research. Increased safety requirements and insufficient knowledge of the thermohydraulic characteristics of these coolants require validation of RANS simulations via information from experiments or DNS. This information can also be applied to improve the RANS turbulence model ifnecessary.In this paper, a series of thermohydraulic DNS simulations of the uniformly heated rod bundle with a relative pitch of 1.4 have been performed to improve RANS. A total of 12 test cases with four different Prandtl numbers of Pr = 0.0043, 0.0154, 0.0324 and 0.71, which correspond to liquid metals and gas, and three Reynolds numbers of Re = 2500, 5000 and 10,000 are considered. The present simulation results show good agreement with the experimental and DNS data. It is revealed that the intermittent flow regime forms at Re = 2500. Furthermore, the turbulent Prandtl number is structured with a maximum of up to y+≈25 and, from y+≈50 onwards, in most cases takes on a constant value. By increasing the Prandtl or Reynolds number, the average value of the turbulent Prandtl number tends to be 1.0. It was also revealed that the turbulent contribution to heat transfer is linear to the Reynolds number for the selected coolant. A new formula for the temperature profile was also obtained, which shows better agreement with DNS data compared to its analogues.As a result, the selected RANS model has been improved (the maximum integral error dropped from 49% to 4%) by the new model of the turbulent Prandtl number and by introducing model coefficients from DNS.

Advancement Toward the Experimental Measurement of the Lead–Bismuth Eutectic Thermal Properties Using Differential Scanning Calorimetry Technique

Abstract The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead–bismuth eutectic (LBE) and lead–lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead–lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in demonstration power plant fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead–lithium, which has significant potential in advanced nuclear systems.

A study of thermophysical and physicochemical characteristic of the KALLA experiment facility

Activities are currently underway across the world to build heavy liquid metal-cooled reactors. For example, the ALFRED and BREST-OD-300 reactor designs use lead coolant, while the MYRRHA and SVBR-100 reactors use lead–bismuth coolant. The high corrosive and erosive activity of the coolant requires control of oxygen concentration and flow velocity. Due to the complex geometry of the reactors cooling circuit flow, numerical simulation methods are used extensively to justify the operation modes for liquid metal coolants. Correct justification of the complex oxygen transport processes in liquid metals requires the detailed physicochemical computational model, which takes into account the major reactions of oxygen with the coolant and the structural materials.This paper presents a physicochemical model, which includes the following processes: erosion, growth, and dissolution of the two-layer oxide film, precipitation and dissolution of metal oxides in the circuit with subsequent deposition in a filter, and inflow of oxygen by the mass exchanger. The physicochemical model was implemented using the model of passive scalar, which allows simulating the impurity transport in the coolant.Capabilities of the presented model are demonstrated based on the results of investigating thermohydraulic and physicochemical processes at the KALLA laboratory experimental facility. The duration of the simulation is 1000 hr.The spatial distributions of metal oxides concentrations, areas of enhanced erosive activity, as well as mass of oxides deposited in the filter and the amount of oxygen entering the circuit have been obtained. The surface distribution of the oxide film thickness on the test facility surfaces contacting the coolant has also been calculated.Accounting for thermocouples and the manufacturing technology made it possible to improve the accuracy of calculating the thermohydraulic characteristics of the coolant flow compared to earlier studies.

Analysis of Factors Influencing Measurement Accuracy for High-Temperature Liquid Metal Flow Sensors in Nuclear Power

Permanent magnet metal flowmeters are implemented to monitor the flow of liquid metal coolant in a nuclear reactor, whose measurement accuracy plays a significant role to ensure the safety and normal operation of the nuclear reactor. According to a theoretical analysis of the permanent magnet metal flowmeter, several factors such as temperature, nonlinear degree, and zero potential will affect the accuracy of measuring the flow. However, for a heavy-caliber permanent magnet metal flowmeter, the influence of the magnetic Reynolds number provides obvious nonlinearity, which affects the measurement accuracy of the flowmeter. Consequently, we use the finite element method to calculate the magnetic field of the flowmeter and analyze the cause of the nonlinearity. Additionally, the influence of nonlinear error is significantly reduced by designing the structure of the flowmeter and the appropriate arrangement of the electrodes.

Based on the temperature correlation principle, the use of a magnetic obstacle to generate pulsations in the flow measurement of a liquid metal coolant

A promising method for determining the flow rate of a liquid metal coolant is the temperature correlation method (TCM) since it does not require calibration. However, being indirect, it has a number of limitations to be carefully studied. A magnetic obstacle is used as a temperature pulsation generator. The paper presents the results of a numerical study on the influence of the size of a magnetic obstacle and its activity ratio, as well as effect of the temperature difference between the liquid metal and the environment, on the performance and accuracy of the TCM. The main criteria influencing the operation of the method are identified, namely the extent and spatial position of the vorticity and mixing zones.

Optimizing the Technological Processes in Cogeneration Power Plants Based on Stirling Engines

The article outlines the advantages of using Stirling engines in small-capacity cogeneration power plants: versatility of the thermodynamic cycle, high efficiency, high environmental cleanliness of working fluids and spent media, the ability to operate on almost all types of conventional fuels and alternative heat sources. The characteristic features of cogeneration power plants based on the Stirling engine are given. The specific features of modeling the thermal cycle circuit and selecting its optimal parameters are considered taking a Stirling liquid piston engine as an example. To solve the practical problem of engine optimization, an economic criterion is proposed as a totality of engine operating parameters. Simulation results are presented, which show that Stirling liquid piston engines can be a promising option as energy converters in installations with a liquid metal coolant.

Fundamental and applied investigations of the liquid-metal cooled fast reactor thermal hydraulics (achieved results and further investigation issues)

The results of experimental and computational-theoretical investigations in the area of hydrodynamics and heat transfer of fast reactors and accelerator-driven system with liquid metal coolants are presented, and the problems and tasks for further investigations are formulated. Physical phenomena, effects, laws and process characteristics occurring in reactor power plants, including a flow path and a fast reactor core: the velocity and temperature fields, the structure and characteristics of turbulent transfer of momentum and energy, the hydraulic resistance of channels and fuel rod assemblies, the collector system hydrodynamics, the vibroacoustics, the heat transfer in channels and fuel rod assemblies with liquid metal cooling and inter-channel exchange are considered and analyzed. The data of experimental investigations are presented on a single-tube model of a large-module steam generator and on a fragmentary thermohydraulic model of a steam generator of a reactor power plant with twisted steam-generating tubes operating at subcritical and supercritical water pressure. The results of investigations of temperature and velocity fields on a small-scale water model of a fast reactor vessel with an integral layout in nominal, transient and emergency operating regimes are demonstrated. The investigation results on model of a fast reactor vessel with an integral layout are demonstrated that the effect of thermogravitational forces leads to temperature stratification with stagnant and recirculating formations, internal waves appear at the stratified boundaries, temperature pulsations, thermal fatigue of structural materials and a decrease in the service life of the equipment. It is shown that the liquid metal boiling process in fuel rod assemblies is formed under the influence of various factors, has a complex structure, is characterized by both stable (nucleate, annular-dispersed) and pulsation (slug) regimes with significant fluctuations of technological parameters, which can cause a crisis of heat transfer. Heat transfer was studied, a cartogram two-phase flow was obtained for liquid metal boiling in fuel rod assemblies, the effect of the surface roughness of fuel rods on liquid metal heat transfer and boiling regimes in fuel rod assemblies was found. The possibility of long-term stable cooling of the core during sodium boiling was shown by using “sodium cavity” over the reactor core. The kinetic and mechanical characteristics of the degradation process of the simulated fuel assembly in fast reactor core during the thermal interaction of the uranium-containing simulators of fuel with static sodium and their dependence on the parameters and design system were determined.Information about the key problems of thermophysical investigations of a high-temperature fast reactor with a sodium coolant for the production of hydrogen and reactors cooled by water at supercritical pressure is given.

Multi-objective optimization of a liquid metal cooled heat sink for electronic cooling applications

Recently, studies on the use of MCHS applied to chips to benefit from the high thermal conductivity of liquid metals have been increasing. This study came up with an optimum choice for MCHS geometric parameters, MCHS material, Gallium-based liquid metal coolant type, and coolant inlet velocity. Total thermal resistance (Rtot), initial investment cost (Ci), pressure drop (ΔP), and maximum operating temperature (Tmax) are selected as performance parameters. I-optimality algorithm used for generation of the Design of Experiment (DoE) based on the Response Surface Method. The optimization parameters and their levels are determined by performing a parametric study. The derived mathematical models for selected performance parameters Rtot, Ci, Tmax, and ΔP, which are also objective functions, are tested by ANOVA with prediction accuracy of 0.983%, 0.997%, 0.886%, and 0.996%, respectively. As a result, the determined optimum values of the channel width, channel height, half of fin width, fin slope angle, and inlet velocity, are 1.16 mm, 3.00 mm, 0.24 mm, 5.96°, and 0.4 m s−1, respectively. In addition, a hybrid structure in which the base of the designed MCHS is aluminum, and the fins are copper, and the GaIn compound for the coolant are calculated as the optimum material options.

Study into the physical chemistry and technology of alkali liquid metal coolants for nuclear and thermonuclear power plants

It is shown that, as the result of developing alkali liquid metal coolants, including sodium, eutectic sodium-potassium alloy, lithium and cesium, the scientific basis has been established for their application in nuclear power. The paper presents data from investigations of thermophysical, neutronic and physicochemical properties and characteristics of various alkali liquid metal coolants, the content of solid-phase and dissolved impurities in coolants, mass transport of impurities in circulation circuits with alkali liquid metal coolants, development of systems for removal of impurities, and control of the content of impurities in alkali liquid metal coolants. Alkali liquid metal coolants are considered as a part of a system that includes a structural material in contact with the coolant, and a gas space that compensates for the thermal expansion of the coolant. The state of the system is defined by the physicochemical properties of the system’s components. And the coolant and the structural materials also represent subsystems consisting of a base material, a coolant and impurities contained both in the material and in the coolant. It has been shown that each alkali liquid metal coolant has its own set of impurities that define its technology. It depends on the physicochemical properties of the solution of the structural material impurities and components in the coolant. Objectives have been formulated for investigating further alkali liquid metal coolants, as stemming from the need to improve the efficiency, environmental friendliness, reliability and safety, and for extending the life of nuclear power plants in operation or under design. Alkali liquid metals are promising candidate materials for being used in thermonuclear power not only as the coolant but also as the tritium breeding medium. These include, first of all, lithium and its eutectic alloy with lead (17 at. % of lithium). The possibility for using lithium or a lithium-lead alloy as a coolant in the blanket of the international thermonuclear power reactor is compared.

Heavy liquid metal cooled fast reactors: peculiarities and development status of the major projects

Fast reactors with heavy liquid metal coolant (lead or eutectic bismuth-lead alloy) are one of the most advanced technologies capable to address the accumulated world nuclear energy issues. This innovative power technology is being developed in Russia, the USA, China and the European Union. Russia is the leader since it has focused on this topic for a number of decades. First concrete started to be poured in June 2021 to form the foundation of the Russian BREST-OD-300 lead cooled reactor scheduled to be started up in 2026. Attention is also given to the development status of the Chinese CLEAR reactor series. A large scope of R&D has been undertaken, and large-scale nonnuclear experimental facilities are under construction. International Euro-US consortiums for the development of the ALFRED, PLFR and MYRRHA reactors do not expect any unsolvable technical issues either and are currently formulating requirements to the test facilities and candidate materials and technologies required for further activities.

Atomistic model of an oxide film in contact with a liquid metal coolant: Defects concentrations and chemical potentials of dissolved Fe–O

The paper considers a prototype atomistic model of an oxide film with a wüstite structure in contact with a liquid metal coolant containing dissolved iron and oxygen. On the basis of direct molecular dynamics modeling with the Tersoff interatomic potential for Fe–O, we check the correspondence between the formation free energies of defects, determined from the temperature dependencies of their equilibrium concentrations, and the formation free energies determined from the calculation of the chemical potentials of dissolved iron and oxygen, the static calculations for defect energies and thermodynamic integration for defect entropies in the oxide lattice. The influence of liquid metal coolant on the chemical potentials of solvated Fe–O is highlighted. These results pave a way for bridging the gap between multiscale models of oxide film formation and ab initio first principles calculations of atomistic models.